ML20138P047
| ML20138P047 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 12/31/1996 |
| From: | Berg W DAIRYLAND POWER COOPERATIVE |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| LAC-13581, NUDOCS 9703040029 | |
| Download: ML20138P047 (145) | |
Text
{{#Wiki_filter:* ylDA/RYLAND hbk[ COOPERAT/VE
- 3200 EAST AVE.SO
- RO. box 817
- LACROSSE, WISCONS (608) 788-4000 FAX NO. (608) 787-1420 WILLIAM L. BERG oznerai uanager February 25,1997 In reply, please refer to LAC-13581 DOCKET NO. 50-409 Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555
SUBJECT:
Dairyland Power Cooperative La Crosse Boiling Water Reactor Possession-Only License No. DPR-45 Annual Radioactive Efiluent Report and Radiological Environmental Monitoring Report
REFERENCES:
(1) NRC Letter, Keppler to Linder, dated August 12,1983, Inspection Report 50-409/83-10 (DRMSP) (2) LACBWR Technical Specifications 6.6.4.2,6.8.1.1.c & d In accor62nce with 10 CFR 50.36a(a)(2), this letter serves to transmit to you the Radioactive EfIluent Report and Radiological Environmental Monitoring Report for the La Crosse Boiling Water Reactor (LACB%%) for 1990. If you have any questions, please contact us. Sincerely, DAIRYLAND POWER COOPERATIVE 9703040029 961231 yJ/21<v M d PDR ADOCK 05000409 R PDR " William L. Berg, General Manager i WLB:LLN:dh \ 030118 Enclosures fh j
,k I cc/ enc: A. Bill Beach, NRC Regional Administrator, Region III l Morton Fairtile, NRC Project Manager l D. Sherman, ANI Library ;
Don Hendrikse, Wisc. Div. of Health ! E5555555555 ,
1 [v^; . i l RADIOACTIVE EFFLUENT REPORT ; AND 4 RADIOLOGICAL ENVIRONMENTAL MONITORING REPORT FOR THE l LA CROSSE BOILING WATER REACTOR (LACBWR) O l (January I to December 31,1996) l 1 DAIRYLAND POWER COOPERATIVE 3200 EAST AVENUE SOUTil LA CROSSE, WI 54602-0817 l l I v l l
l l TABLE OF CONTENTS , O SECTION A: RADIOACTIVE EFFLUENT REPORT Section lille Pagg Introduction 1 1.0 Regulatory Limits 2 2.0 Efiluent Release Concentration Limit 4 3.0 Average Energy 4 4.0 Analytical Methods 4 5.0 Batch Releases 6 6.0 Abnormal Releases 7 ; 7.0 Estimated Total Analytical Error 7 8.0 Offsite Dose Calculation Summary and Conclusions 13 9.0 Offsite Dose Calculation Manual (ODCM) Changes 14 10.0 Process Control Program (PCP) Changes 16 O l 1 SECTION B: RADIOLOGICAL ENVIRONMENTAL MONITORING REPORT - Section Title P_ age Introduction 1 1.0 Sample Collection 3 2.0 Results of the 1996 Radio-Environmental Monitoring Survey 3 3.0 Conclusions 6 4.0 Interlaboratory Comparison Program Results 6 I 1 O i I
I l 1 TABLE OF CONTENTS I TABLES
- SECTION A
- RADIOACTIVE EFFLUENT REPORT l Table No. Iills Eagg 1'
1A Emuent and Waste Disposal - Gaseous Emuents 8 l Summation of All Releases IB Emuent and Waste Disposal - Gaseous Emuents 9 Elevated Release 2A Emuent and Waste Disposal - Liquid Emuents 10 Summation of All Releases 2B Emuent and Waste Disposal - Liquid Emuents 11 3 Solid Waste and Irradiated Fuel Shipments 12 SECTION B: RADIOLOGICAL ENVIRONMENTAL MONITORING REPORT l O V i Table No. Title Page I l 5 Sample Frequency and Analysis of Radio-Environmental 11 l Samples 6 Permanent Environmental Monitoring Station Locations 12 ) 7 Environmental TLD Locations 13 4 8 Radio-Environmental Samples Collected 14 Jann.try-December 1996 9 Quarterly Thermoluminescent Dosimeter Dose Measurements 15 in the LACBWR Vicinity 10 Weekly Gross Beta Air Particulates in the LACBWR Vicinity 16 11 Air Particulate Composite Results 19 l 12 Results of Analysis of Mississippi River Water in the Vicinity of 25 LACBWR O l ii l
i TABLE OF CONTENTS l TABLFf -(cont'd) SECTION B: RADIOLOGICAL ENVIRONMENTAL MONITORING REPORT Table No. Tills Page l l l 13 Results of Analysis of Mississippi River Sediment in the 31 Vicinity of LACBWR l 14 Results of Analysis of Milk Samples in the Vicinity of 32 LACBWR 15 Fish Sample Activity in the Vicinity of LACBWR 33 16 Vegetation Sample Activity in the Vicinity of LACBWR 35 FIGURES O NA Tjlig Page FIGURE 1 LACBWR Prcperty Map 7 FIGURE 2 Permanent Environmental Monitoring Station Locations 8 FIGURE 3 LACBWR Environmental Dose Assessment Locations 9 FIGURE 4 LACBWR Environmental Dose Assessment Locations 10 O iii
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O I 1 l l SECTION A RADIOACTIVE EFFLUENT O REPORT l l l l 1 i I l 4 ) k !. O
l 1
.O !
i INTRODTE' TION: i l l The La Crosse Boiling Water Reactor (IACBWR), also known as Genoa Station No. 2, is l located on the east bank of the Mississippi River near Genoa, Vernon County, Wisconsin. The plant was designed and constructed by the Allis-Chalmers Manufacturing Company. It was completedin 1967 and had a generation capacity of50 MW(165 MWQ. The reactor is owned l by DairylandPower Cooperative (DPC). 1 The reactor went criticalin July 1967 andfirst contributed electricity to DPC's system in April 1968. After completingfullpower tests in August 1969, the plant operated between 60% and 100%fullpower, with the exception ofplant shutdownsfor maintenance and repair. in April of1987 plant operation was ceased The reactor ispresently defueledand in a SAFS70R mode. In August of1987 apossession-only license was received. In accordance with IACBWR TechnicalSpecifications 6.8.1.1.d and in compliance with j 10 CFR 50.36a(a)( j, C:is document is the Radioactive Effluent Reportfor the period.h ,i 1 through December 31,1996. 1 4 !O l [ __ _- .-
1 EFFLUENT AND WASTE DISPOSAL REPORT O (SupplementalInformation) FACILITY: La Crosse Boilln_ g Water Reactor LICENSEE: Dairvland Power Cooperative DOCKET NO. 50-409 l 1.0 REGULATORY LIMITS
- a. Oa. sing; Emuent Release Limits:
LACBWR's Technical Specifications for gaseous emuent releases of radioactive material limit the release rates of the sum of the individual radionuclides, so that the dose rates to members of the public beyond the Emuent Release Boundary do not exceed l 500 mrem / year to the whole body,3000 mrem / year to the skin from noble gases, and i 1500 mrem / year to a critical organ from H-3 and paniculates with half-lives greater than 8 days. 1 O Also, in accordance with 10 CFR 50, Appendix I, the Technical Specifications for 1 gaseous emuent radioactive material limit the air dose to a member of the public from ! l noble gases in areas beyond the Emuent Release Boundary to less than 5 mrad gamma and 10 mrad beta per calendar quarter, and less than 10 mrad gamma and 20 mrad beta per
- calendar year. The dose limits from H-3 and particulates with half- lives greater than j 8 days are less than 7.5 mrem per calendar quarter, and less than 15 mrem per calendar year to any organ.
l Cumulative dose contributions from gaseous emuent releases are determined in l accordance with the LACBWR Offsite Dose Calculations Manual. O m._ . _ _ _ _ _ _ _ . _ . _ .._ _ _ _ _ _ _ _ _ .__ l I
~ EFFLUENT AND WASTE DISPOSAL REPORT -(cont'd)
- b. Liquid Emuent Release Limits: ,
LACBWR's Technical Specification limits for liquid emuent releases are those concentrations specified in 10 CFR 20 Appendix B, Table 2, Column 2, for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration is limited to a total activity concentration of 6 E-4 pCi/ml. For alpha emitting radionuclides, the concentration is limited to a total activity concentration of 4.9 E-9 pCi/ml, based upon an actual alpha emitting radionuclide analysisperformed on a representative water sample. The values reported in tables 2A and 2B, Liquid Emuents, are based on dilution with the combination of LACBWR and Genoa Station No. 3 condenser cooling water flow prior to discharge to the Mississippi River. No credit is taken for further dilution in the mixing zone of the Mississippi River. Also, in acco'rdance with 10 CFR 50, Appendix I, the dose commitment to a - member of the public from radioactive materials released in liquid emuents to areas beyond the Emuent Release Boundary are limited to less than 1.5 mrem whole body and' 5.0 mrem organ dose per calendar que.rter, and less than 3.0 mrem whole body and 10 mrem organ dose per calendar year via the critical ingestion pathway. Cumulative quarterly and annual dose contributions from liquid emuev w .ases are determined for the adult fish ingestion pathway in accordance with the LACBWR , Offsite Dose Calculation Manual. V l EFFLUENT AND WASTE DISPOSAL REPORT-(cont'd) ; i O c. Solid Radioactive Waste
- All solid radioactive wastes are handled in accordance w% a Process Control ;
Program as defined by LACBWR procedures in order to assure that all applicable transportation and burial site disposal requirements are met. , t 2.0 EFFLUENT RELEASE CONCENTRATION LIMIT k l The Efiluent Release Concentration used to calculate permissible release rates are obtained i from 10 CFR 20, Appendix B, Table 2, Columns 1 and 2. In addition, the following values are used: Tritium in Water = 1 E-3 pCi/ml. Tritiumin Air = 1 E-7 pCi/cc. 3,0 AVERAGE ENERGY The release rate limits for LACBWR are not based on average energy. 4.0 ANALYTICAL METHODS
- a. Liquid Efiluents Liquid eflludnt measurements for gross radioactivity are performed by HPGe gamma isotopic analysis of a representative sample from each tank discharged. In addition, each batch discharged tank is analyzed for alpha and tritium activity j concentration. A composite sample is created by collecting representative aliquots from each tank batch discharged and is analyzed quanerly for Iron-55 and Strontium-90 by a contractor.
O l
EFFLUENT AND WASTE DISPOSAL REPORT -(cont'd) O
- b. Airborne Particulates.
1 Airborne particulate releases are determined by HPGe gamma isotopic analysis. This analysis is performed by analyzing a glass fiber filter paper taken from the stack monitor which continuously isokinetically samples and monitors the stack efIluent. This filter is changed and analyzed on an approximate weekly basis and analyzed within 7 days after removal. This filter is also analyzed for alpha activity. A quarterly composite of these filters is sent to a contractor for Sr-90 analysis. < l
- c. Radiciodines Since the plant shutdown in April 1987, the I-131/I-133 have decayed completely to stable elements. Amendment 66 to the LACBWR Technical Specifications, Table .
5.7.2.2, no longer requires monitoring for iodine. O d. ' Fission and Activation Gasss The concentration of radioactivity (pCi/cc)in gaseous releases from the stack is continuously monitored by two stack monitors, which are in-line monitors. These gas concentrations (pCi/cc) are corrected for pressure loss in the sampling system and averaged by the monitors microprocessor. The resu'its are used along with the stack flow rate to obtain the daily gaseous release from the plant. Since the plant shutdown in April 1987, gaseous releases have been immeasurable. All fission gases except Kr-85 have l decayed completely to stable e:cments. . t O l
1 l EFFLUENT AND WASTE DISPOS AL REPORT - (cont'd) ! l l i. i O) L
- e. Tritium Tritium releases are determined by taking a grab sample of the stack atmosphere at j the emuent of the stack monitor. Tntium, as tritiated water, is removed from the sample stream by condensation, using a cold trap. The condensed water vapor is then distilled and the distillate is analyzed for H-3 concentration, pCi/cc, by internal liquid scintillation spectrophotometry and the results are expressed in terms of tritium release rates. The ;
tritium grab samples are obtained on at least a once/ month basis unless the upper reactor l cavity is flooded, at which time the sampling frequency is increased to at least once per 7 days. 5.0 B ATCH RELEASES
- a. Airborne All airborne emuent releases at LACBWR are from a single Continuous-Elevated Release Point.
- b. Liquid All liquid emuent releases at LACBWR are batch releases. This is summarized as follows:
(1) Number of Batch Releases: 28 (2) Total Time Period for Batch Releases: 221.2 hours (3) Maximum Time Period for a Batch Release: 19.0 hours (4) Average Time Period for a Batch Release: 7.90 hours (5) Minimum Time Period for a Batch Release: 2.3 hours l (6) Average Stream Flow Rate During Periods of Release of Emuent into a Flowing Stream: 61,900 ff/sec n
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EFFLUENT AND WASTE DISPOSAL REPORT - (cont'd) O 6.0 ABNORMAL RELEASES There were no abnormal releases of radioactivity in plant efiluents as summarized as follows:
- a. liquici (1) Number ofReleases: None l
(2) Total Activity Released. N/A
- b. Gaseous (1) Number ofReleases: None I
(2) Total Activity Released: N/A 1 l I I 7.0 ESTIMATED TOTAL ANALYTICAL ERROR l ! The reported analytical results contain the following estimated errors: Counting Error
- 1 Standard Deviation Sampling Volume Error
- 5%.
1 The lower limits of detection (LLD) are expressed in terms of a 4.66 o as defined in Technical Specifications. l LO
EFFLUENT AND WASTE DISPOSAL REPORT - (cont'd) TABLE 1A P) ( EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT 1996 GASEOUS EFFLUENTS - SUMMATION GF ALL RELEASES UNIT QTR QTR QTR QTR TOTAL A. FISSION & ACTIVATION GASES
- 1. TOTAL RELEASE Ci 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00
- 2. AVERAGE RELEASE RATE FOR pCi/ 0.00E+00 0.00E+00 0.00E+00 0.00E+00 PERIOD Sec H. 10 DINE I-131 - No longer analyzed for.
C. PARTICULATFS 1
- 1. PARTICULATES WITil Ci 8.13E-9 0.00E+00 0.00E+00 1.28E-6 1.29E-6 IIALF-LIVES > 8 DAYS
- 2. AVERAGE RELEASE RATE FOR pCi/ 1.03E-9 0.00E400 0.00E+00 1.67E-7 PERIOD Sec U 3. GROSS ALPilA RADIOACTIVITY Ci 3.90E-8 0.00E+00 0.00E+00 0.00E+00 3.90E-8 D. TRITIUM
- 1. TOTAL RELEASE Ci L92E-2 4.12E-2 2.52E-2 1.36E-2 9.92E-2
- 2. AVERAGE RELEASE RATE FOR Ci/ 2.44E-3 5.24E-3 3.17E-3 1.71E-3 ,
PERIOD Sec E. PERCENTAGE OF (APPENDIX I) TECilNICAL SPECIFICATION LIMITS QTR QTR QTR QTR YEARLY
- 1. NOBLE GAS RELEASE GAMMA % 0.00E+00 0.00E400 0.00E+00 0.00E+00 0.00E+00 BETA % 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00
- 2. 11-3 AND ALL RADIONUCLIDES IN l
PARTICULATE FORM WITil HALF-LIVES GREATER TilAN 8 DAYS ! lilGilEST ORGAN % l .06E-4 2.22E-4 1.36E-4 1.31E-4 2.96E-4 i EFFLUENT AND WASTE DISPOSAL REPORT -(cont'd) O TABLE 1B EFFLUENT AND WASTE DISPOSAL ANNUAI, REPORT 1996 GASEOUS EFFLUENTS - ELEVATED RELEASE CONTINUOS MODE UNIT QTR QTR QTR QTR TOTAI. NUCLIDES RELEASED
- 1. FISSION GASES KRYPTON-85 Ci 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00
- 2. IODINE I-131 - No longer analyzed for.
- 3. PARTICULATES STRONTIUM-90 Ci 8.13E-9 0.00E+00 0.00E+00 0.00E400 8.13E-9 CESIUM-134 Ci 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E400 CESIUM 137 Ci 0.00E+00 0.002+00 0.00E+00 1.28E4 1.28E-6 COBALT 40 Ci 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Ci Ci Ci Ci Ci Ci Ci 8.19E-9 0.00E+00 0.00E+00 1.28E4 1.1;J4 TOTALS O
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EFFLUENT AND WASTE DISPOSAL REPORT -(cont'd) TABLE 2A. . EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT 1996 L10UID EFFI,UENTS- SUMMATION OF ALL REi, EASES , UNIT QTR l QTR l QTR QTR l TOTAL l
.A. FISSION & ACTIVATION PRODUCTS ,
- 1. TOTAL RELEASE (NOTINCL. Ci 4.4 IE-4 1.28E-2 6.32E-3 9.26E 3 2.88E-2 TRITIUM, GASES, ALPHA)
- 2. AVERAGE DILUTCD CONCEN- pCi/ 5.79E-10 3.82E-9 3.28E-9 1. l lE-8 TRATION DURING PERIOD ml
- 11. TRITIUM
'l. TOTAL RELEASE Ci 5.93E-3 3.55E-2 2.58E-2 9.83 E-3 7.70E-2
- 2. AVERAGE DILUTED CONCEN- pCi/ 7.78E-9 1.06E-8 1.34E-8 1.18E-8 TRATION DURING PERIOD ml
, DISSOLVED AND ENTRAINED GASES O 1. TOTAL RELEASE Ci 0.00 0.00 0.00 0.00 0.00 b 2. AVERAGE DILUTED CONCEN- Ci/ 0.00 0.00 0.00 0.00 TRATION DURING PERIOD ml D. GROSS ALPIIA RADIOACTIVITY
- 1. TOTAL RELEASE Ci 0.00E+00 2.45E-5 2.21E-5 9.84E-6 5.73E-5 E. VOLUME OF WASTE RELEASED Liters 4.80E+4 1,82E+5 8.18E+4 6.88E+4 3.80E+5 (PRIOR TO DILUTION)
F.- VOLUME OF DILUTION WATER Liters 7.62E+8 3.34E+9 1.93E+9 8.32E+ 8 6.86E+9 USED DURING PERIOD G. PERCENTAGE OF (APPENDIX I) TECllNICAL SPECIFICATION LIMITS FOR LIQUID P.ELEASES QTR QTR QTR QTR YEARLY llIGilEST ORGAN % 0.12 1.22 1.I 1 0.85 1.65 WilOLE BODY % 0.25 2.60 2.34 1.82 3.50 l
EFFLUENT AND WASTE DISPOSAL REPORT -(cont'd) ) I
,q TABLE 2B V '
EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT 1996 LIOUID EFFLUENTS l 1 NUCLIDES RELEASED UNIT QTR QTR QTR QTR MANGANESE-54 Ci 0.00E+00 0.00E+00 0.00E+00 0.00E+00 l IRON-55 Ci 1.23E-5 3.03E-4 2.43E-4 1.96E-3 ! COBALT-60 Ci 9.31 E-5 8.77E-3 1.83E-3 5.78E-3 ; STRONTIUM-90 Ci 2.56E-5 1.06E-4 7.51E-5 4.74E-5 l CESIUM-134 Ci 0.00E+00 0.00E+00 0.00E+00 0.00E+00 CESIUM-137 Ci 3.10E-4 3.61 E-3 4.17E-3 1.48E-3
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I TOTAL FOR PERIOD Ci 4.41E-4 1.28E-2 6.32E-3 9.26E-3 l (ABOVE) i l l KRYPTON-85 Ci 0.00E+00 0.00E+00 0.00E+00 0.00E+00 l l 4 l 1 \ O l i
--,w e .M---b-- - 4 14'- p3AAAJ.- "A.4;+--6 >4A4--a -. --*s-*a A 45 A ap .-4 b ta-4-.- ,.,- -_ -A A-EFFLUENT AND WASTE DISPOSAL REPORT - (cont'd)
TABLE 3 EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT - 1996 , SOLID WASTE AND IRRADIATED FUEL SIIIPMENTS A. SOLID WASTE SillPPED OFFSITE FOR DISPOSAL (Not Irradiated Fuel) 6-MONTil 6-MONTII
- 1. TYPE OF WASTE UNIT PERIOD PERIOD 1 TOTAL 4 ,
- a. SPENT RESINS, FILTER SLUDGES, m' 0 0 0 EVAPORATOR DOTTOMS, ETC. Ci 0 0 0 !
- b. DRY COMPRESSIBLE WASTE, m' 108.75 0 108.75 ,
CONTAMINATED EQUIPMENT, ETC. Ci 0.31 0 0.31
- c. IRRADIATED COMPONENTS, CONTROL m' O 0 0 RODS. ETC- Ci 0 0 0
- d. OTIIER(DESCRIBE) m' 0 0 0 Ci o o o
- 2. ESTIMATE OF MAJOR NUCLIDE 6-MONTH 6-MONTH COMPOSITION (BY TYPE OF WASTE) PERCENT PERIOD L PERIOD-Co-60 35.3 0.109 Cs-137 8.9 0.028 Fe-55 28.3 0.088 Ni-59 4.9 0.015 ,
Ni-63 22.2 0.069
- 3. SOLID WASTE DISPOSITION NO. OF StilPMENTS MODE OF TRANSPORTATION DESTINATION 2 Sole Use American Ecology (Oak Iddge, TN)
H. IRRADIATED FUEL SIIIPMENTS (DISPOSITION) { DESTINATION 110. OF SIIIPMENTS MODE OF TRANSPORTATION - NONE 12
EFFLUENT AND WASTE DISPOSAL REPORT -(cont'd)
- 8. O OFFSITE DOSE CALCULATIONS
SUMMARY
AND CONCLUSIONS: l
- a. Gaseous Efiluent Releases The maximum quarterly offsite gamma dose due to noble gases was 0.00 mrad.
The cumulative 1996 annual offsite gamma dose due to noble gases was 0.00 mrad. The maximum quarterly offsite beta dose due to noble gases was 0.00 mrad. The cumulative 1996 annual offsite beta dose due to noble gases was 0.00 mrad. The maximum quarterly offsite dose to any organ from the release of H-3 and all radionuclides in particulate form with half-lives greater than 8 days was approximately 1.66 E-5 mrem. The cumulative 1996 annual maximum organ dose from these radionuclides was also approximately 4.41 E-5 mrem. The highest historical annual average X/Q equal to 1.82 E-6 sec/m' for the period 1985-1987 for the worst case offsite receptor location, in accordance with the ODCM, i h' was used to calculate these offsite dose values.
- b. Liquid Effluent Releases The maximum quarterly organ dose from liquid releases was approximately 6.11 E-2 mrem. The maxiiaum cumulative 1996 annual organ dose was approximately 1.65 E-1 mrem. The maximum quarterly total body dose for liquid releases was approximately 3.90 E-2 mrem, and the cumulative 1996 annual total body dose was approximately 1.05 E-1 mrem.
4
- c. Conclusion )
l All calculated offsite doses were below Technical Specification limits. l \ i LO 13 l
EFFLUENT AND WASTE DISPOSAL REPORT -(cont'd) O 9.0 OFFSITE DOSE CALCULATION MANUAL (ODCM) CHANGES The ODCM has been completely rewritten and reorganized in 1996 in anticipation of approval of LACBWR's reduction in Technical Specifications due to SAFSTOR. The present Technical Specification requirements were included into the ODCM to ensure present requirements are not dropped or forgotten when the new approved Technical Specifications are received. Also, a Radioactive Emuent Control Program (RECP) has been added to conform to the guidance of 10 CFR 50.36a. The RECP will contain all the efiluent release criteria presently in Technical Specifications. The Radiological Environmental Monitoring Program (REMP) has been expanded to ensure the present environmental Technical Specification requirements are contained in it. The ODCM is being included in this report for your review, as required by Technical Specifications. O i l Changes to ODCM
- 1. ,The following definitions now in Technical Specifications have been added to the ODCM.
a) Channel calibration b) Channel check c) Channel Functional Test d) Efiluent Release Boundary e) Exclusion Area , f) Member of the Public g) Operable-Operability h) Source Check O 14
l EFFLUENT AND WASTE DISPOSAL REPORT -(cont'd) I w) II. The following Technical Specification requirements (Amendment 66) were placed into the RECP: , Technical Specification Reauirement RECP Location 4.2.4 Section 3.3 4/5.2.4 Section 3.3 4.7.1 Section 3.2 4.7.2 Section 3.3 ! 4/5.7.1 Section 3.2 4/5.7.2 Section 3.3 4.7.4 Section 3.4 4/5 7.4 Section 3.4 (3 V ; 111. The following Technical Specification requirements (Amendment 66) were placed into the REMP: Technical Specification Requireme_nt REMP Location 4.8.1 Section 4.5 5.8.1.1 Section 4.2 5.8.1.2 Section 4.4 Table 4.8.1-1 Section 4.2, Table 4.1 Table 4.8.1-2 Section 4.3, Table 4.2 Bases (page 4/5-61) Section 4.1
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_ . . _ . . . . . __- ....__._.._._.___.____._...____._._.__-.._.._._m. . l EFFLUENT AND WASTE DISPOSAL REPORT -(cont'd) ; i 10.0 PROCESS CONTROL PR6 GRAM (PCP) CHANGES l i In 1996 the PCP was rewritten. The requirements of the PCP have not been changed but i l: . i rather reorganized to provide easier readability of this document. Due to submitted changes to ) l l LACBWR Technical Specifications, the reference Technical Specification number has been j i l removed as a reference, but the Technical Specification requirements remain. This was done in l l anticipation of NRC approval of LACBWR's Technical Specification submittal. A safety analysis l l l for the PCP changes is included in the change. The PCP is being included in this report for your
.)
review, as required by Technical Specifications. l lO 1 I I l i l0 l 16 l
i J a 1 i O . I i i 1 f i i l. SECTION B l l I i 1 1 1 i ANNUAL i RADIOLOGICAL ! ENVIRONMENTAL MONITORING , j O RerOnT i
- January 1 - December 31,1996 i
i i e i l ) i 'l I l I , a i, lO 1 i i l 4
l l I INTRODUCTION: 4 The RadiologicalEnvironmentalMonitoring (REM) Program is conducted to comply with the requirements of TechnicalSpecifications and in accordance with 10 CFR 50 Appendix 1. l
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The REM Program provides measurements ofradiation and ofradioactive materials in those exposure pathways andfor those radionuclides which couldpotentially lead to radiation doses to l l Members of the Public resultingfrom plant effluents. Environmentalsamples are taken withm i l the surrounding areas of the plant and in selected control or background locations. I The monitoringprogram at the IACBWRfacility includes monitoring ofliquid and gaseous releasesfrom the plant, as well as environmentalsamples ofsurface air, river water, river sediment, milk, fish, andpenetrating radiation. 7he REMprogram theory supplements the Radioactive Effluent analyses by verifying that the measurable concentrations ofradioactive materials andlevels ofradiation are not higher than expected on the basis of the effluent measurements and modeling of the environmental exposure pathways using the methodology of the Offsite Dose Calculation Manual (ODCM). An Interlaboratory Comparison Program is provided to ensure that independent checks l on the precision andaccuracy of the measurements ofradioactive materialin environmental swnples are performed. i O I I
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4 I l RADIOLOGICAL ENVIRONMENTAL MONITORING REPORT - (cont'd) IA CBlVR was removedfrom the power generation rolls in April 1987 when DPC shut down the reactor andplacedIACBWR in SAFSTOR. Since that time, the amount ofradioactive i b materialreleased to the environment has decreasedsigmpcantly. Amendment 64 to the i 1 IACBIVR Technical Specifications was issued in December of1988 and eliminated the need to anal):e environmental samplesfor I-131 due to the stabili:ation ofalliodine at IA CBWR. 4 Amendment 66 to the Technical Specifications was receivedin August of1991 andsigmpcantly l reduced the sampling andanalysis requirements ofIA CBWR's environmentalprogram. 4 i IACBWR's in-house procedures were modipedin February of1992. At that time IACBWR environmentalprogram was reduced to match that listed in Technical Specipcation Q Amendment 66. l l 4 ! i ! I ' j
- i 4
e N i 4 .a
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n - -
l RADIOLOGICAL ENVIRONMENTAL MONITORING REPORT - (cont'd)
) 'y.9 SAMPLE COLLECTION ]
O Envircnmental samples are collected from the area surrounding LACBWR at the l frequencies outlined in the Technical Specifications. A series of figures and tables are included in this report to better show LACBWR's environmental program. ! FIGUlG 1 This map includes the plant boundary, roads, other generation plants, and the relationship of the plant to the nearest local community. 4 FIGURE 2 This map shows the location of LACBWR's permanent environmental l monitoring stations. I These maps show the location of LACBWR's TLDs. FIGURES 3&4 i TABLE 5 This table shows the sampling frequency of the various environmental samples and the analyses performed on these samples ! TABLE 6 This table shows the permanent monitoring stations used in LACBWR's i environmental program. ( TABLE 7 This table shows the TLD locations. TABLE 8 This table shows the number of various samples collected and analyzed during
- 1996.
1 2.0 RESULTS OF TIIE 1996 RADIO-ENVIRONMENTAL MONITORING SURVEYS t i i During 1996, activity levels in the local environment were normal, indicating no significant ! plant attributed radioactivity. 4 2.1 PENETRATING RADIATION l The environmental penetrating radiation dose is measured by thermoluminescent j i dosimeters consisting of four lithium fluoride (LiF) chips. These TLD's are changed on a i quarterly basis and are sent to an outside contractor for reading. The TLD results for 1996 are shown on Table 9. lO 3-
RADIOLOGICAL ENVIRONMENTAL MONITORING REPORT - (cont'd) 2.2 AIR PARTICULATE Air samples are collected continuously from various sites (see Table 6) around LACBWR. An air sampler is also located 18 miles north of the plant in La Crosse, Wisconsin, to act as a control station. Particulate air samples are collected at the rate of approximately 30-60 Ipm with a Gelman Air Sampler. The air filter consists of a glass fiber filter with an associated pore size of approximately 0.45 pm. The particulate filters are analyzed weekly for gross beta activity with an internal proportional counter, and the monthly particulate composites are gamma analyzed for individual isotopic concentration. TABLE 10 This table shows the weekly gross beta gamma activity concentration from the air particulate filters. O T^nLE ii Thistebieshewsthecemgeeiteairparticeieteisetenicenairsis. Comparison between the control station at La Crosse and the other stations near LACBWR indicate that there was no significant plant attributable airborne particulate activity. 2.3 RIVER WATER River water is collected monthly. River water samples above, at, and below the plant site are collected and are gamma analyzed for isotopic concentration. The river water gamma isotopic analysis results are shown in Table 12. The results indicate that there is insignificant plant-attributable radionuclides in the river water. O
~ _
RADIOLOGICAL ENVIRONMENTAL MONITORING REPORT - (cont'd) p 2.4 SEDIMENT SAMPLES U Sediment samples were collected twice per year above, at, and below the plant outfall. These samples were gamma analyzed and these results appear on Table 13. They indicated that small amounts of plant attributed radionuclides have accumulated in river sediments near the outfall. 2.5 MILK SAMPLES Milk samples were collected monthly during the grazing season (May through September) from two dairy farms in the vicinity of LACBWR (see Table 6) and gamma analyzed. These samples are listed in Table 14. There has been no significant accumulation of plant attributed radionuclides in milk in the vicinity of LACBWR. 2.6 FISH Fi h samples were collected quarterly above and below the plant discharge. The results of gamma spectral analysis of edible portions of fish samples appear in Table 15. There has been no significant accumulation of plant attributed radionuclides in fish in the vicinity of LACBWR. 2.7 VEGETATION Vegetation samples were collected from local gardens at the time of harvest. The i results of the gamma spectral analysis of the vegetation samples appear on Table 16. There has been no significant accumulation of plant attributed radionuclides in the vegetation. O RADIOLOGICAL ENVIRONMENTAL MONITORING REPORT - (cont'd)
3.0 CONCLUSION
S All environmental samples collected and analyzed during 1996 exhibited no significant contribution from LACBWR. 4.0 INTERLAHORATORY COMPARISON PROGRAM RESULTS , During 1996, interlaboratory comparison samples were obtained from an outside contractor. The equipment used to analyze the environmental samples was tested against the contractors' results. The following is the result of this comparison.
-LACBWRi ' CONTRACTOR---
ANALYSIS 'RESULTS (pCi/cc)': 'RESULTS (pCi/cc) GROSS BETA 2.63 E-3 1.86 E-3 GROSS ALPHA 8.83 E-4 8.49 E-4 O Tn>Tiua 7.22 8-3 7.32 8-3 CS-137 6.35 E-3 6.46 E-3 CO-60 3.72 E-3 3.68 .I-3 e U t f f VILL A GE Of l ..? E!?.A . i i
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3 i 4 i FIGURE 3 - LACBWR ENVIRONMENTAL DOSE ASSESSMENT LOCATIONS
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- 10 -
4
O O O TABLE 5 SAMPLE FREOUENCY AND ANALYSIS OF RADIO-ENVIRONMENTAL SAMPLES i SAMPLE FREOUENCY ANALYSIS PERFORMED TLD (LiF) Dosimeters Quarterly Dose in mrem Particulate Air Glass Fiber Filters Weekly Gross Beta and Gamma Spectroscopy of Composites Monthly (HPGe-MCA) , t Milk Monthly during grazing season Gamma Spectroscopy Sediment Twice per year Gamma Spectroscopy Fish Quanerly Gamma Spectroscopy I River Monthly Gamma isotopic analysis and tritium (Liquid Scintillation Analyzer) Vegetation At time of harvest Gamma Spectroscopy i t
d
- TABLE 6 PERMANENT ENVIRONMENTAI, MONITORING STATION LOCATIONS (Refer to Figure 2)
LOCATI.ON:: _ . _ . ..
.; AIR-SNO;; '? LOCATION? ;LSAMPLE :. '1 MILK-::-
i 1 Pedretti Farm x l f 2 P. Malin Farm x 4 I 3 Radio Tower x 4 Dam No. 8 x 5 Trailer Court x 0 1 l 6 Crib House x 7 Main Office x 0
; RADIOLOGICAL ENVIRONMENTAL MONITORING REPORT - (cont'd) j TABLE 7 I
ENVIRONMENTAL TLD LOCATIONS I
- LOCATION ,
i ?NOT LOCATIONL. j 1 LA CROSSE MAIN OFFICE AIR SAMPLER BOX 2 DAM #8 AIR SAMPLER BOX 3 RADIO TOWER BUILDING AT AIR SAMPLER 4 TRAILER COURT AIR SAMPLER BOX 5 CRIBHOUSE AIR SAMPLER BOX 5 G-3 CONTROL ROOM
- 6 I
7 SW GATEPOST AT END OF G-3 DIKE 8 ON FENCE N. SIDE OF FISHERMAN'S ROAD 9 SITE ENTRANCE GUARD AREA 1 10 ON FENCE AT NE CORNER OF THE SWITCHYARD 11 ON N. SITE AREA FENCE GATE 12 G-1 CRIBHOUSE 13 ON MOORING WALKWAY WEST OF LACBWR #2 WAREHOUSE 14 G-3 COAL UNLOADING CRANE 15 POWER POLE ON BLUFF SIDE EAST OF PLANT l l 16 RESTRICTED AREA FENCE N. SIDE j 17 RESTRICTED AREA FENCE E. SIDE 18 RESTRICTED AREA FENCE S. SIDE 19 RESTRICTED AREA FENCE SW CORNER 20 RESTRICTED AREA FENCE W. SIDE 21 RESTRICTED AREA FENCE NW CORNER l l l
RADIOLOGICAL ENVIRONMENTAL MONITORING REPORT - (cont'd) O TABLE 8 RADIO-ENVIRONMENTAL SAMPLES COLLECTED JANUARY-DECEMBER 1996 NUMBER OF TYPE OF SAMPLE SAMPLES Penetrating Radiation (TLD's) 83 Air Particulate 257 River Water (includes Ellluent Split w/ State of Wisc.) 36 Sediment 6 Milk 12 Fish 8 Vegetation 2 O 14
l l RADIOLOGICAL ENVIRONMENTAL MONITORING REPORT - (cont'd) TABLE 9 OUARTERLY THERMOLUMINESCENT DOSIMETER DOSE MEASUREMENTS IN THE LACBWR VICINITY 1996
) STATION 1st QUARTER 2nd QUARTER. 3rd QUARTER . J4th QUARTER 1 1 NO.- mrem ' - mremi mrem . ' mrem .
I 17.2
- 5.4 12.4
- 6.7 14.6
- 4.6 13.4
- 5.2 2 18.6
- 5.4 15.0
- 4.7 18.6
- 5.8 18.8 0.9 3 19.8 5.0 13.4
- 6.'i 16.8
- 4.3 13.8
- 4.3 4 22.4 2.3 14.0
- 2.8 18.2 6.7 19.0
- 1.4 5 29.6
- 3.3 15.0 8.6 21.2
- 9.3 22.4 4.6 6 18.6 1.8 12.2 3.8 MISSING 13.6 1.8 i 7 21.2
- 5.2 15.6
- 6.9 21.8
- 7.1 20.6 t 1.1 8 20.4 + 6.6 13.4 7.6 21.4 2.5 19.0 + 1.4 9 25.8
- 2.6 18.6
- 4.6 21.8 3.8 17.4 6.1 10 22.4 3.0 16.8 3.0 16.2
- 7.0 15.8
- 4.3 11 23.6
- 5.0 17.8 2.6 20.2 1.7 18.8 7.9 l 12 24.0 7.1 17.4 6.4 22.2
- 6.1 19.2 4.8 !
l3 29.6
- 2.3 18.8 9.0 20.0
- 6.8 21.8 9.1 j 14 19.4 + 3.6 12.0
- 2.8 15.2
- 0.9 14.0
- 6.5 15 28.0
- 2.0 22.0 5.5 25.8
- 3.3 25.4
- 4.1 16 62.0 14.6 39.2
- 17.6 44.2 11.4 59.0
- 19.5 ,
1 17 128.2
- 20.2 73.4
- 51.7 60.8 11.3 61.8 25.7 18 53.4
- 19.3 44.8
- 16.5 56.8
- 9.5 51.8
- 14.7 19 49.2 8.3 31.0 6.8 35.6 + 3.3 33.8
- 7.4 20 41.6 6.1 27.8
- 6.5 28.6 + 3.3 26.8
- 5.9 21 54.8 13.4 34.0 22.4 35.0
- 11.5 40.4 7.6 Station #1 (La Crosse Main Office) locatedapproximately 16 miles north ofIACBWR is considered the Control TLD.
O O O TABLE 10 WEEKLY GROSS BETA AIR PARTICULATES IN THE LACBWR VICINITY , (Reporting Level = 10 times Control Value) COLLECTION-- LACBWR PLANT ~ TRAILER COURT 1 DAM. #8y RADIO T_OWER? iLA CROSSE? DATE :. pCi/m - 2 fpCi/m'? fpCi/m" ThCi/nf/
' ECONTROIR 1/9/96 .037 i.004 .042 i .004 .052 t.004 .040 i.004 .042 i .003 1/16/96 .041 .004 .020 1.002 .046 i .003 .041 i .004 .043 i .003 i'
1/23/96 .041 i.004 .026 i .003 .048 i .003 .041 .003 .039 .003 1/30/96 .028 1.003 .025 i.003 .032 i .003 .015 t.002 .032 i .003 2/6/96 .047 i.004 .046 i.004 .058 i.004 .021 i .003 .054 i .003 2/13/96 .028 1.003 .023 i .003 .033 i.003 .023 i.003 .021 i.002 2/20/96 .030 1.003 .026 .003 .032 i.003 .029 i.003 No Sample Pump OOC 2/27/96 .025 .003 .024 i .003 .025 i .003 .024 i.003 .027 1 .003 . 3/5/96 .029 1.003 .028 i.003 .037 i .003 .035 .003 .029 .003 3/12/96 .027 t .003 .023 i.003 .055 i.008 .013 i.002 .031 i.003 l 3/19/96 .025 .003 .024 i .003 No Sample .034 i .003 .027 .003 Pump OOC ! 3/26/96 .025 i.003 .022 i.003 .020 .003 .023 i .003 .022 i .002 4/1/96 .019 1.003 .025 i .004 .018 1.003 .032 i.003 .022 i.003 4/9/96 .028 i.003 .027 i.003 .031 .003 .034 i .003 .031 .003 - 4/16/96 .022 .003 .021 1.003 .032 i.004 .026 1.003 .026 i.003 l 4/23/96 .021 i.002 .016 1.002 .022 i.003 .023 i.003 .022 .002 4/30/96 .020 1.003 .017 .003 .023 i .003 .022 i.003 .022 .003 I
, - e N5h TABLE 10 WEEKLY GROSS BETA AIR PARTICULATES IN THE LACBWR VICINITY (Reporting Level = 10 times Control Value) COLLECTION LACBWR PLANT TRAILER COURT DAM #8 : RADIO TOWERE LA CROSSE - DATE pCi/m' pCi/m' pCi/m' ' pCi/m's . CONTROL. 5/7/96 .018 .003 .014 i .003 .022 1.003 .022 i.003 .017 i .002 5/14/96 .016 i.003 .019 i.003 .026 1.003 .025 1.003 .018 1.002 5/21/96 .019 .003 .018 .003 .024 i.003 .026 1.003 No Sample Pump OOC 5/28/96 .021 i .003 .019 i .002 .015 1 .002 .025 i.003 .020 i.002 6/4/96 .021 i.003 .018 .003 .017 i .002 .020 i.003 .018 1.002 6/11/96 .015 .002 .015 i.002 .015 i .002 .018 .003 No Sample Pump OOC 6/18/96 .017 .003 .019 i.003 .019 i .002 .023 i .003 .022 i.002 6/25/96 .014 .002 .013 i.002 .017 .003 .015 i.002 .014 i.002 7/2/96 .023 .003 .023 i.003 .033 i .003 .023 t.003 .022 i .002 7/9/96 .019 .002 .021 1.002 .025 i .003 .020 .003 .020 i.002 7/16/96 .022 .002 .021 1.003 .029 i.003 .027 .005 .026 t .003 7/23/96 .025 .003 .020 i .003 .039 .004 .036 i.003 .027 .003 7/30/96 .019 .003 .018 i.002 .020 .003 .026 i.003 .019 .002 8/6/96 .024 .003 .022 i.003 .024 i.003 .034 i.004 .019 .002 8/12/96 .018 .003 .019 .003 .020 i.003 .027 i.003 .020 t .003 8/19/96 .030 i.003 .027 .003 .030 .003 .035 .004 .030 .003 8/27/96 .023 .003 .022 .003 .025 .003 .036 .004 .026 .002
O O O TABLE 10 WEEKLY GROSS BETA AIR PARTICULATES IN THE LACBWR VICINITY (Reporting Level = 10 times Control Value) COLLECTION LACBWR PLANT TRAILER COURT DAM #80 RADIO TOWERS : LA CROSSEi DATE pCi/m'. pCi/m' :pCi/m'? - pCi/m' ; CONTROL 9/3/96 .039 1.004 .035 .003 .046 i .004 .052 i .004 .044 i .003 9/9/96 .063 .005 .060 i.004 .068 i .005 .075 .006 .068 i.004 9/17/96 .020 .002 .019 .002 .027 i.003 .029 i.003 .021 i.002 9/24/96 .028 .003 .027 i.003 .036 i .003 .036 i .004 .029 i .003 I 10/1/96 .025 .003 .021 i .003 .023 .003 .028 i .004 .033 i.007 , 10/8/96 .029 1.003 .029 i.003 .036 i.004 .059 .002 .029 1.003 10/15/96 .030 i.003 .028 i.003 .037 i.004 No Sample .031 .003 Pump OOC 10/22/96 .028 i.004 .026 i.003 .031 i .003 .029 i .003 .030 .003 10/29/96 .026 .003 .029 i.003 .029 i.003 .033 i .003 .030 i .003 11/5/96 .026 .005 .027 1.003 .035 i.004 .031 i.003 .030 i .003 11/12/96 .019 1.003 .021 i .003 .029 .003 .027 i .003 .024 i .003 11/19/96 .029 .003 .027 i.003 .033 i .004 .025 .003 .027 i.003 11/26/96 .018 .003 .017 i .003 .022 1.003 .016 i .003 .023 .003 12/3/96 .048 1 .004 .042 .004 .046 i .004 .037 i.003 .048 i.003 12/10/96 .028 i .003 .030 i.003 .028 1.003 .021 i .003 .016 1.002 12/17/96 .030 .003 .029 i .003 .029 i .004 .023 .003 .032 1 .003 12/23/96 .040 .004 .036 i .004 .052 1.005 .028 .003 .044 i .004 12/30/96 .043 .004 .040 i.003 .055 i .005 .020 .003 .046 i .003
l O _ O . O TABLE 11 AIR PARTICULATE COMPOSITE RESULTS (Concentrations in pCi/m') LOCATION RADIO TOWER LACBWR TRAILER COURT DAM NO. 8 LA CROSSE START DATE 1-02-96 1-02-96 1-02-96 1-02-96 1-02-%
- END DATE l-30-96 1-30-96 1-30-96 1-30-96 1-30-96 ISOTOPES /RL* .
Cs-134/10 <l .90E-3 <l .98E-3 <2.59E-3 <l.76E-3 <l .47E-3 ' Cs-137/20 <2.04E-3 <2.12E-3 <2.59E-3 <1.66E-3 <l .67E-3 [ *RL = REPORTING LEVEL LOCATION RADIO TOWER LACBWR TRAILER COURT DAM NO. 8 LA CROSSE START IlATE 1-30-96 1-30-96 1-30-96 1-30-% l-30-96 END DATE 2-27-96 2-27-96 2-27-96 2-27-96 2-27 l ISOTOPES /RL* 1' Cs-134/10 <l .87E-3 <2.10E-3 <2.83E-3 <l .97E-3 <2.1 IE-3 Cs-137/20 <l .94 E-3 <2.10E-3 <2.13 E-3 <l.88E-3 <2.08E-3 a
*RL = REPORTING LEVEL
A m n kj U TABLE 11 AIR PARTICULATE COMPOSITE RESULTS ' (Concentrations in pCi/m') LOCATION RADIO TOWER LACBWR TRAILER COURT DAM NO. 8 LA CROSSE START DATE 2-27-96 2-27-96 2-27-96 2-27-96 2-27-96 END DATE 4-01-96 4-01-96 4-01-96 4-01-96 4-01-96 ISOTOPES /RL* Cs-134/10 <l .51 E-3 < l .71 E-3 <l .52E-3 <2.69E-3 <l .36E-3 Cs-137/20 <l .56E-3 <l .74 E-3 <l .71E-3 <2.56E-3 <l .31E-3
- RL = REPORTING LEVEL LOCATION RADIO TOWER LACBWR TRAILER COURT DAM NO. 8 LA CROSSE START DATE 4-01-96 4-01-96 4-01-96 4-01-96 4-01-96 END DATE 4-30-96 4-30-96 4-30-96 4-30-96 4-30-96 ISOTOPES /RL*
Cs-134/10 <l .74E-3 <2.08E-3 <2.01E-3 <2.06E-3 <1.45 E-3 Cs-137/20 <2.03 E-3 <2.00E-3 <l .94E-3 <2.00E-3 <l .49E-3
*RL = REPORTING LEVEL
O O O TABLE 11 AIR PARTICULATE COMPOSITE RESULTS (Concentrations in pCi/m') LOCATION RADIO TOWER LACBWR TRAILER COURT DAM NO. 8 LA CROSSE START DATE 4-30-96 4-30-96 4-30-96 4-30-96 4-30-96 END DATE 5-28-96 5-28-96 5-28-96 5-28-96 5-28-96 ISOTOPES /RL* Cs-134/10 <2.08E-3 <1.92E-3 <2.12E-3 <2.03 E-3 <2.1 SE-3 Cs-137/20 <2.04E-4 <l .98E-3 <2.12E-3 <2.23 E-3 <2.19E-3
- RL = REPORTING LEVEL LOCATION RADIO TOWER LACBWR TRAILER COURT DAM NO. 8 LA CROSSE START DATE 5-28-96 5-28-96 5-28-96 5-28-96 5-28-96 END DATE 7-2-96 7-2-96 7-2-96 7-2-96 7-2-96 ISOTOPES /RL*
Cs-134/10 <l .62E-3 <l .61 E-3 <l .60E-3 <l .62E-3 <l .63 E-3 Cs-137/20 <l .74 E-3 <l .61.E-3 <l.73E-3 <l .49E-3 <l .62E-3
- RL = REPORTING LEVEL
_ _ _ _ _ _ _ _ _ _ _ _____ _____. _ _ _ _ _ _ _ ____________j
]
TABLE 11 AIR PARTICULATE COMPOSITE RESULTS ' (Concentrations in pCi/m') LOCATION RADIO TOWER LACBWR TRAILER COURT DAM NO 8 LA CROSSE START DATE 7-02-96 7-02-96 7-02-96 7-02-96 7-02-96 END DATE 7-30-96 7-30-96 7-30-96 7-30-96 7-30-96 ISOTOPES /RL* Cs-134/10 <2.4SE-3 <2.01 E-3 <2.06E-3 <2.17E-3 <1.67E-3 Cs-137/20 <2.57E-3 <l .97E-3 <2.16E-3 <2.08E-3 <l .63 E-3
*RL = REPORTING LEVEL LOCATION RADIO TOWER LACBWR TRAILER COURT DAM NO. 8 LA CROSSE START DATE 7-30-96 7-30-96 7-30-96 7-30-96 7-30-96 END DATE 9-3-96 9-3-96 9-3-96 9-3-96 9-3-96 ISOTOPES /RL*
Cs-134/10 <2.00E-3 <l .57E-3 <l .60E-3 <l .57E-3 <1.32E-3 Cs-137/20 < l .96E-3 <l .60E-3 <l .67E-3 <l .66E-3 <1.26E-3
*RL = REPORTING LEVEL
O O O TABLE 11 AIR PARTICULATE COMPOSITE RESULTS (Concentrations in pCi/m') LOCATION RADIO TOWER LACBWR TRAILER COURT DAM NO. 8 LA CROSSE START DATE 9-03-96 9-03-96 9-03-96 9-03-96 9-03-96 END DATE 10-1-96 10-1-96 10-1-96 10-1-96 10-1-96 ISOTOPES /RL* Cs-134/10 <2.68E-3 <l .95E-3 <2.08E-3 <2.12E-3 <l .90E-3 Cs-137/20 <2.54 E-3 <l .98E-3 <2.07E-3 <2.17E-3 <2.15E-3
- RL = REPORTING LEVEL LOCATION RADIO TOWER LACBWR TRAILER COURT DAM NO. 8 LA CROSSE START DATE 10-01-96 10-01-96 10-01-96 10-01-96 10-01-96 END DATE I1-05-96 11-05-96 11-05-96 11-05-96 11-05-96 ISOTOPES /RL*
Cs-134/10 <2.84E-3 < l .86E-3 <l .60E-3 <l.88E-3 <l .22E-3 Cs-137/20 <2.50E-3 <2.03 E-3 <l .68 E-3 <l .87E-3 <l .31E-3
- RL = REPORTING LEVEL
( O O O TABLE 11 AIR PARTICULATE COMPOSITE RESULTS (Concentrations in pCi/m') LOCATION RADIO TOWER LACBWR TRAILER COURT DAM NO. 8 LA CROSSE - START DATE 11-05-96 11-05-96 11-05-96 11-05-96 11-05-96 END DATE 12-03-96 12-03-96 12-03-96 12-03-96 12-03-96 ISOTOPES /RL* Cs-134/10 <2.22E-3 <2.42E-3 <2.28E-3 <2.74E-3 <l .81E-3 Cs-137/20 <2.20E-3 <2.63 E-3 <2.22E-3 <2.63 E-3 <l .76E-3
*RL = REPORTING LEVEL LOCATION RADIO TOWER LACBWR TRAILER COURT DAM NO. 8 LA CROSSE START DATE 12-03-96 12-03-96 12-03-96 12-03-96 12-03-96 END DATE 12-30-96 12-30-96 12-30-96 12-30-96 12-30-96 ISOTOPES /RL*
Cs-134/10 <2.19E-3 <2.43 E-3 <2.27E-3 <2.83E-3 <l.82E-3 Cs-137/20 <2.33 E-3 3.45E-3 6.88E-4 <2.26E-3 <3.01 E-3 <1.82E-3 . I l
*RL = REPORTING LEVEL t
_ _ _ _ _ _ _ _ - . .__ . _ _ _ . _ _ _ _ _ _ _ _ - _ _ - . - . - _ _ _ - - _ _ _ _ _ - _ _ - _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ _ - - - _ _ _ _ - - - _ _ -__ ___ a
, ~ . _ . _
O O O TABLE 12 RESULTS OF ANALYSIS OF MISSISSIPPI RIVER WATER IN THE VICINITY OF LACBWR (Concentrations in pCi/ Liter) SAMPLE #1 -SAMPLE #2 - SAMPLE #3:: SAMPLE #1 - SAMPLE #2 SAMPLE #3' COLLECTION DATE: 1-9-96 1-9-96 .1-9-96 2-6-96' 2-6-96 12-6-96: SAMPLE LOCATION: DAM 8 OUTFALL VICTORYJ - DAM 8 : OUTFALL.- VICTORY - ISOTOPES /RL
- H-3/ 20000 824 1320 982 i 325 792 320 <341 <341 <341 Mn-54/1000 <4.96 <5.21 <5.19 <5.25 <5.27 <5.20 Co-60/300 <5.48 <5.25 <5.06 <5.21 <5.23 <5.03 Zn-65/300 <12.0 <11.7 <11.7 <11.7 <11.9 <11.2 Cs-134/30 <5.43 <5.35 <5.32 <5.51 <5.22 <5.70 Cs-137/50 <5.58 <5.21 <5.84 <5.61 <5.09 <5.58 i
- RL = REPORTING LEVEL '
O O O TABLE 12 RESULTS OF ANALYSIS OF MISSISSIPPI RIVER WATER IN TIIE VICINITY OF LACBWR (Concentrations in pCi/ Liter) SAMPLE #1 - SAMPLE #2 ' SAMPLE #3 :- SAMPLE #1 i SAMPLE #2 ' SAMPLE #3 -- COLLECTION DATE: 3-5-96 3-5-96 3-5-96 '4-1-96: 14-1-96 l l 4-1 SAMPLE LOCATION: DAM 8- _ OUTFALL : VICTORY ; ;-DAM Si ;:OUTFALL :.VICTORYe ISOTOPES /RL
- 11-3/ 20000 463 i 329 <340 <340 663 i 328 567 i 326 682 i 328 Mn-54/1000 <5.16 <4.95 <5.27 <5.22 <5.19 <5.19 Co-60/300 <4.93 <5.26 <4.56 <5.18 <5.25 <5.14 Zn-65/300 <10.6 <11.6 <11.0 <11.4 <11.3 <10.7 Cs-134/30 <4.87 <5.30 <5.27 <5.47 <5.23 <5.62 Cs-137/50 <5.49 <5.44 <5.36 <5.28 <5.63 <5.26
- RL = REPORTING LEVEL
O O O TABLE 12 RESULTS OF ANALYSIS OF MISSISSIPPI RIVER WATER IN THE VICINITY OF LACBWR (Concentrations in pCi/ Liter) SAMPLE #1- ' SAMPLE #2 - --SAMPLE #3 c f SAMPLE #1 -- SAMPLE #2 - SAMPLE #3 - COLLECTION DATE: 5-7-96 5-7-961 5-7-96 6-4-96 '- : 6-4 6-4-% SAMPLE LOCATION: ~ DAM 8 : OUTFALL':: ? VICTORY? DAM 8 . 1OUTFALLi s. VICTORY::.: ISOTOPES /RL
- H-3/ 20000 529 i 326 515 i 326 1492 355 631 342 655 342 708 i 342 Mn-54/1000 <5.13 <4.87 <5.33 <5.12 <5.02 <5.07 Co-60/300 <5.03 <4.70 <4.95 <5.20 <5.06 <4.85 Zn-65/300 <11.3 <11.6 <10.3 <11.5 <11.1 <10.7 Cs-134/30 <5.17 <5.39 <5.33 <5.44 <5.62 <5.11 Cs-137/50 <5.48 <5.35 <5.10 <5.57 <5.49 <5.68
- RL = REPORTING LEVEL O O O TABLE 12 RESULTS OF ANALYSIS OF MISSISSIPPI RIVER WATER IN THE VICINITY OF LACBWR (Concentrations in pCi/ Liter)
SAMPLE #1 SAMPLE #2. SAMPLE #3 SAMPLE #1 - .. SAMPLE #2. SAMPLE #3: COLLECTION DATE: 7-2-96 7-2-96 7-2-96' 8-12-96 .8-13-96 .8-13-96 SAMPLE LOCATION: . DAM 8 .OUTFALL VICTORY- / DAM 8 OUTFAIL . VICTORY; ISOTOPES /RL
- H-3/ 20000 407 i 325 4861327 525 i 328 <365 <365 <365 Mn-54/1000 <5.32 <4.94 <5.21 <5.39 <4.96 <5.37 Co-60/300 <5.30 <5.12 <5.16 <5.04 <5.00 <5.39 Zn-65/300 <11.3 <10.8 <10.9 <11.4 <12.0 <11.5 Cs-134/30 <5.36 <5.26 <4.87 <5.37 <5.43 <5.37 Cs-137/50 <5.23 <5.3 5 <5.43 <5.65 <5.47 <5.10
- RL = REPORTING LEVEL ;
O O O TABLE 12 RESULTS OF ANALYSIS OF MISSISSIPPI RIVER WATER IN THE VICINITY OF LACBWR (Concentrations in pCi/ Liter) SAMPLE #1 - SAMPLE #2- SAMPLE #3 - SAMPLE #1 SAMPLE #2 SAh&LE #3 ' COLLECTION DATE: 9-10-96 9-10-96 9-10-96 10-8-96' ..-10-8-96 ~ 10-8-96~ SAMPLE LOCATION: DAM 8 ' OUTFALL VICTORY . - DAM 8 - OUTFALI2 i VICTORYi l ISOTOPES /RL
- H-3/ 20000 <362 <362 <362 <235 374 i229 261i226 Mn-54/1000 <5.24 <4.87 <5.33 <5.30 <5.01 <5.26 Co-60/300 <5.11 <5.06 <5.3 8 <4.80 <4.83 <5.26 Zn-65/300 <11.3 <10.9 <10.9 <10.6 <10.6 <11.1 t Cs-134/30 <5.08 <5.59 <5.11 <5.16 <5.00 .
<5.13 Cs-137/50 <5.28 <5.48 <5.43 <5.15 <5.13 <5.13 s
- RL = REPORTING LEVEL ,
O O O TABLE 12 RESULTS OF ANALYSIS OF MISSISSIPPI RIVER WATER IN THE VICINITY OF LACBWR (Concentrations in pCi/ Liter) SAMPLE #1 SAMPLE #2 SAMPLE #3 ' SAMPLE #1 - ' SAMPLE #2 - SAMPLE #3 COLLECTION DATE: 11-12-96 11-12-96. 11-12-96 .12-10-96' .12-10-96 12-10-96-; SAMPLE LOCATION:~ DAM 8- 'OUTFALL- : VICTORY ~ ~ DAM 8 (OUTFALL - VICTORY ' ISOTOPES /RL
- H-3/ 20000 <235 <235 <235 359i229 <235 <235 Mn-54/1000 <4.76 <4.84 <4.58 <4.72 <4.45 <4.71 Co-60/300 <5.06 <5.08 <5.30 <5.26 <4.86 <5.01 Zn-65/300 <11.2 <11.3 <10.8 <10.8 <10.6 <10.6 Cs-134/30 <5.28 <5.44 <5.34 <5.16 <5.09 <5.27 Cs-137/50 <5.31 <5.73 <5.55 <4.68 <4.60 <4.68 4
- RL = REPORTING LEVEL '
O O O TABLE 13 RESULTS OF ANALYSIS OF MISSISSIPPI RIVER SEDIMENT IN THE VICINITY OF LACBWR (Concentrationin pCi/Kg) (Reporting Level = 10 times Upstream Value) COLLECTION DATE SAMPLE # 1 SAMPLE # 2 SAMPLE # 3 SAMPLE # _1 SAMPLE # 2 SAMPLE # 3 SAMPLE LOCATION 5-29-96 5-29-96 5-29-96 8-21-96 8-21-96 8-21-96 DISCHARGE UPSTREAM DOWNSTREAM DISCHARGE - UPSTREAM _ DOWNSTREAM ISOTOPES Co-60 214 i 5.59 <6.56 <10.3 36818.9 <6.15 <8.42 Cs-134 <8.47 <6.68 <11.9 <9.48 <6.46 <9.58 Cs-137 774 28.0 11.5 i 1.77 167
- 7.1 1558 55 <6.11 15.0 2.1 I
l l TABLE 14 RESULTS OF ANALYSIS OF MILK SAMPLES IN TIIE VICINITY OF LACBWR (Concentrations in pCi/ Liter) l
? SAMPLE ' COLLECTION ' -ISOTOPES /RL * ~
LOCATION - .:DATE:: lCs-134/60T Cs-137/60:: P. MALIN 1-9-96 <5.66 <5.93 PEDRETTI 2-13-96 <5.87 <6.15 P. MALIN 3-12-96 <5.94 <5.63 PEDRETTI 4-9-96 <5.46 <5.77 P. MALIN 5-14-96 <5.7 <5.78 PEDRETTI 6-11-96 <5.72 <5.94 P. MALIN 7-9-96 <5.72 <6.07 PEDRETTI 8-13-96 <5.68 <5.86 i P. MALIN 9-9-96 <5.79 <5.98 PEDRETTI 10-8-96 <5.91 <5.99 P. MALIN 11-8-96 <5.97 <5.87 PEDRETTI 12-10-96 <5.44 <5.24
- RL = REPORTING LEVEL O
O O O TABLE 15 FISil SAMPLE ACTIVITY IN THE VICINITY OF LACBWR (Concentrations in pCi/Kg) COLLECTION DATE: SAMPLE #1 SAMPLE #2 SAMPLE #1 SAMPLE #2 SAMPLE #1 SAMPLE #2 FISH SPECIES: 3-19-96 3-29-96 5-29-96 6-20-96 8-20-96 8-20-96 CARP SHEEPHEAD CARP SHEEPHEAD CARP PIKE ISOTOPES /RL* Mn-54/3E4 <12.5 <8.14 <7.15 <9.80 <6.08 <7.82 Co-60/IE4 < 13.6 <9.33 <8.48 <10.3 <6.75 <8.20 Zn-65/2E4 <31.0 <21.3 <18.4 <25.6 <15.6 <20.3 Cs-134/IE3 <l 2.9 <8.59 <8.04 <10.7 <6.62 <8.08 Cs-137/2E3 <13.6 <9.26 <8.10 <9.64 <6.61 <7.99
- RL = REPORTING LEVEL
O O O TABLE 15 FISII SAMPLE ACTIVITY IN Tile VICINITY OF LACBWR (Concentrations in pCi/Kg) COLLECTION DATE: SAMPLE #1 SAMPLE #2 SAMPLE #1 SAMPLE #2 SAMPLE #1 SAMPLE #2
+
FISH SPECIES: 10-24-96 10-24-96 CARP PIKE ISOTOPES /RL' Mn-54/ 3E4 <7.23 <8 05 . Co-60/ IE4 <7.79 <8.63
~
Zn-65/ 2E4 <l9.0 <20.1 Cs-134/ IE3 <7.52 <8.3 8 Cs-137/ 2E3 <7.46 <8.88
*RL = REPORTING LEVEL
_ _ . _ . _ _ _ _ . ____.____._-__.______.__________.__________m _ _ _ _ _ _ . _ _ _ _ _ _ - _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___
O # E L P M A S E R L P W M B A C S A L F O # Y E T L I P N M I C A S V )g I E K
/
H C i 6 T o E 1 N i n L - E I s P L Y no M 5 OB 3 T i t A _ A I a S _ T V t r T cn I e 2 C n Y E A o 6
# 9- 1 T D O V V I
A N E C 1 E 3 C- D U N T O ( L L2 I B M O P P -5 A C A L F M M A A A S S N O I T 1 Y T E D A 6
# 9- 0 O V I
V A N T D U N IT O 1 E E 3 C-L2 B M O G P -5 A C A F E M L A V A S E T A D S E N P O O I T T C O S E I L L O O C /
O LA CROSSE BOILING WATER REACTOR (LACBWR) OFFSITE DOSE CALCULATION MANUAL l Prepared by 4 Health Physics Review: m Y////'/6 V f I' / Date O oueiiiv ^eeurence Revie-: 54 c/- << - - rc !
. Date ORC Approved: /W #/ N A 9/
/~I bat 6 l
March 1996 Revision 4 Dairyland Power Cooperative 3200 East Avenue South La Crosse, WI 54602-0817 O t
TABLE OF CONTENTS e w. O J 1. INTRODUCTION ... . .. . ....... ............... ..... .. .... ....... 1
-.y ,/ 1.1 Purpose .. .. . .. . .. . .... 1 g <
1.2 Definitions ....... . .. ... . ... ....... . ...... .. 1 s E 8 2. OFFSITE DOSE CALCULATIONS B, 1 2.1 Compliance with the Limitations for Liquid Effluent Releases 5 (
$ 2.2 Compliance with the Limitations for Gaseous Effluent Releases .... . 12 8 -
ib oE%> 3. RADIOACTIVE EFFLUENT CONTROL PROGRAM u 3.1 Program Requirements .. .... ... ... .......... ... .. ........ 37
- s N
b ( 3.2 Liquid Effluents ... ............ ....... .. ...... ... 3.2.1 Sampling and Analysis . . . . . . . . . . . . . . . . .... . 38 38 3.2.2 Liquid Effluent Release Limitation .... . 40 9s g 3.2.3 Liquid Effluent Instrumentation . .. .
. .. . 42 E< 3.3 Gaseous Effluents . .. . .. . ...... . ... . . .. . 45
.L3T ' 3.3.1 Containment Building Ventilation . . . . . . . . . . . ... .. ..... 45 i 3.3.2 Stack Effluent Sampling and Analyses .. .... ... .. . . 45 i f 3.3.3 Stack Effluent Release Limitation . . .. .... . 47 j
I j 3.3.4 Instrumentation .. .. . .. . . .. . . 52
- 4. RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM e 8 w a o g 4.1 Program Requirements .. .. . . . . 59 g 4.2 REMP Description . . 60
?> z y 8 4.3 REMP Lower Limits of Detection . . 63
$ Ej 4.3.1 Calculation of Lower Limits of Detection 64 g z 1
v J 66 4.4 Interlaboratory Comparison Program e >. . . . E $ 4.5 Reporting Requirements .. . . . 66 p G l ODCM i Rev. 4 l
TABLE OF CONTENTS -(cont'd)
%' LIST OF TABLES e
16 o - 2.1 Infant Dose Factors Pa (Inhalation) for H-3 and Particulate s Gaseous Release Monitor Alarm Setpoint Determinations . . . . . . . . . . . . . . 31 e R ' E 3.1 Radioactive Liquid Waste Sampling and Analysis Requirements g for Batch Releases . . . . . . . . . . . . . . .. ................ ..... .......... . 39 l\w p 5 3.2 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance
$ Requirements . . . . . . . . . . . . . ...... .. ..... . . .... . .. 43 ? ?J 0
R 4 3.3 Radioactive Gaseous Waste Sampling and Analysis . . . . . . . . . . . . . . . . . . . 46 S $ 3.4 Radioactive Gaseous Effluent Monitoring Instrumentation . . . . . . . . . . . . . . . 54 8 tv O, s 3.5 Redioective Geseove effiuent uonitorino instrumentetion S rveiiience Requirements . . . . .... .......... .. . ......... .. . .. ..... . 55 5 E - 8 . f 4.1 c. Radiological Environmental Monitoring Program . . . . . . . . . . . . . . . . . . . . . . 61 E
$ ,4.2 Environmental Sample Analyses Lower Limits Values (LLD) . . . . . . . . . . . . 65 ==
3 $ 4.3 Reporting Levels for Radioactivity Concentrations in Environmental 8 $ Samples . . . . .. ... . . .. . . . .. .. .. .. 68 n 5 3 5 x w s 15 Z v a e >- EE
! 5 ODCM ii Rev. 4
TABLE OF CONTENTS -(cont'd) ; { LIST OF FIGURES 2 o , 2.1 Liquid Release Monitor Alarm Setpoint Determination ..... . .. . 11 1 E. 2.2 Noble Gas (Kr-85) Release Monitor Alarm Setpoint Calculations .. . . 32 2.3 Noble Gas (Kr-85) Release Monitor Alarm Setpoint Summary 33 h .. . .. s
).
lE %a 2.4 H-3 and Particulate Gaseous Release Monitor Alarm Setpoint y Determination .. .. .. ... . . . . ... . . .... . . . .... 34 i b 0E( 2.5 H-3 and Particulate Gaseous Helease Monitor Alarm Setpoint Summary 35 4 1 j g 2.6 Air Dose Commitment to a Member of the Public from Noble Gas
>- (Kr-85) Release . . ... .... ... .. . .... ... .. .... 36 3: Ds l 54 I &. 8 \
gM i ik E LIST OF DIAGRAMS l [ 1 sm 1 y i 4 1.1 Site Map, including Effluent Release Boundary ... .. . . . 4 N e l 3 E E z
.e o E $
5 E v i 1 5 gr-ODCM iii Rev.4
I T
1.0 INTRODUCTION
n %h 5 %. v 1.1 Purpose _l : r
& The OFFSITE DOSE CALCULATION MANUAL (ODCM) contains the 5
l methodology and parameters used in (1) the calculation of offsite doses resulting from O 'y g ; radioactive gaseous and liquid effluents from LACBWR, (2) the calculation of gaseous g. g ( and liquid effluent monitoring AlarmTTrip Setpoints, and (3) the conduct of the O i V Environmental Radiological Monitoring Program. The ODCM also contains the b Radioactive Effluent Controls and Radiological Environmental Monitoring Programs. 4 S $ 5* 1.2 Definitions g h CHANNEL CAllBRATION h A CHANNEL CAllBRATION shall be the adjustment, as necessary, of the s }b channel output such that it responds with the necessary range and accuracy to known f i values of the parameter whicn the channel monitors. The CHANNEL CAllBRATION g shall encompass the entire channel including the sensor and alarm and/or trip w Ds U functions, and shall include the CHANNEL FUNCTIONAL TEST. The CHANNEL N E CAllBRATION may be performed by any series of sequential, overlapping or total 3 8 g $ channel steps such that the entire channel is calibrated. E E O . 3 N aE
& 5 .
O-- ODCM 1 Rev. 4
1 i CHANNEL CHECK 4 p_ g og A CHANNEL CHECK shall be the qualitative assessment of channel behavior l i- during operation by observation. This determination shall include, where possible, (? e J comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter. l
& CHANNEL FUNCTIONAL TEST x
E b A CHANNEL FUNCTIONAL TEST shall be: p a. Analog channels - the injection of a simulated signal into the channel as N 5 (v close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions and channel failure trips. iO 1 E b. Bistable channels - the injection of a real or simulated signal into the
.0 -
sensor to veriiy OPERABILITY including alarm and/or trip functions. f' g l 2 I L
- EFFLUENT RELEASE BOUNDARY l m 1
$ $ The Dairyland Power Cooperative property line within the 1109 ft. (338m) radius l R 1 EXCLUSION AREA is the EFFLUENT RELEASE BOUNDARY. (See Diagram 1.1.)
di
? z
$ 8
$N u
g
& 5 O
ODCM 2 Rev. 4 i
4 l ! __ EXCLUSION AREA ) % {Q o y The EXCLUSION AREA is defined as the area within an 1109 ft. (338m) radius i> from the centerline of the Containment Building. This was the area established per 10 e '
& CFR 100 as the EXCLUSION AREA for plant siting and operation.
d : E 4 O h, MEMBER OF THE PUBLIC E D E E MEMBER OF THE PUBLIC shall mean an individual in a CONTROLLED or 5Y 8 UNRESTRICTED AREA. However, an individual is not a MEMBER OF THE PUBLIC oD during any period in which the individual receives an occupational dose. S $ 5 k OPERABLE-OPERABILITY m b g } A system, subsystem, train, component or device shall be OPERABLE or have L g' ' OPERABILITY when it is capable of performing its specified function g s g necessary attendant instrumentation, controls, a normal or an emergency electrical 5
- g power source, cooling or seal water, lubrication or other auxiliary equipment that are !
I s required for the system, subsystem, train, component or device to perform its e j $ function (s) are also capable of performing their related support function (s). R l { 1 E SOURCE CHECK l b> 6 e !
$o $ A SOURCE CHECK shall be the qualitative assessment of channel response u a
$ g when the channel sensor is exposed to a radioactive source.
em j
& 5 :
ODCM 3 Rev. 4
Diagram 1.1
~~
( ' g SITE MAP INCLUDING EFFLUENT RELEASE BOUNDARY w o . J
/ -
u I o a v / 8 o et4ETEOROLOGtCAL TOWER seus .- 4' [ . 4 o 4oo- .- -1 { l r d
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8 i , d%%f o Ef% ,. / EEUNEE
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a
, sc m u
" "5' 4 o.ri .tc cretusiou / L AC BWR- -
1 Ioo@t
- SouuoArx ,
e z -',s
+ 300 -
a ee - sG
~'
/eFFLUEUT "l % '
(m
' 6
\ . ' RELEASE souuoAny ; '
i _t.iouio Errtutui - oiscutac.c id , l!. N
's t cincutarma g '
wAnt oiscuisar i 1 a. TA : I E . 8 I si .s I' A (
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e 8 * [ I a g g p' e [, . hg
- 555 m - a r u l, *
- 9 '
f y f 5 i b lI 3 8 ( l' t s i l; 1 IE u J - i in fl li s s ii
/ ;I ODCM 4 Rev. 4
O-, 2.0 OFFSITE DOSE CALCULATIONS g' .. O' __ 2.1 Compliance with the Limitations for Liauld Effluent Releases
? ?
e e a) To assure compliance with the limitations of Section 3.2.2.a, Radioactive Effluent I 4 .
/ !j a Control Program (RECP), radioactivity monitor alarm setpoints are calculated for each 6
B radioactive liquid effluent line monitor as a function of the maximum effluent flow rate 3 and the minimum dilution flow rate. The following equation is used to calculate 8 5 h setpoints: J b ! sC (2.1)
% k( F f) i t o g y where:
N C = the effluent concentration limit implementing 10 CFR 20 for g g LACBWR, in Ci/ml. 9
.L3d. s a = the setpoint (in CPS above background) of the radioactivity monitor }
o.
$ measuring the radioactivity concentration in the effluent line prior to $. r dilution and subsequent release; the setpoint, which is inversely 3 $ proportional to the volumetric flow of the effluent line (f) and 0 0 "
proportional to the volumetric flow of the dilution stream plus the fs 8 z effluent stream (F + f), represents a value which, if exceeded, could y d result in concentrations exceeding the limits of 10 CFR 20. E Z 1 d
,a g k = the conversion factor, cps per pCi/ml, for the liquid waste effluent a 5 ,
monitor based upon most recent calibration of the monitor. ODCM 5 Rev. 4 l
i 1 i
\
l l ! l l
-- f = the effluent line volumetric flow setpoint as measured at the radiation Q. '
monitor location, in gallons per minute. O
. 1 g F = the dilution stream (LACBWR & Genoa Station No. 3 (G-3] l e
E Condenser Cooling Water) volumetric flow, in gallons per minute. 4 LI E i e 1 0h Since f << F, Equation 2.1 it satisfied when the following discharge line radioactivity l 3 ! 5 1 g g monitor setpoint is met: E \ 4 a s; kCF (2.2) g( f oS 9 E O t' Calculation of instantaneous Allowable Release Rates v l
,o o g7 N LACBWR's liquid radwaste is released in batches. In order to assess the i %
@ k" required radioactive liquid effluent line monitor setpoint, a, the following step-by-step 8
2A
' method for obtaining data will be performed. The form presented in Figure 2.1 may be
[ E . i used as a worksheet for these calculations. The alarm setpoint calculation may be I 4 N performed on an annual basis if the setpoint is determined to be sufficiently { $ conservative so as to prevent exceeding 0.5 MPC at the discharge point where MPC is 3 the isotope weighted effluent concentration release limit for a typical LACBWR waste E E z batch based on 10CFR20, appendix B, Table 2, Col. 2 values. 2 O E $ 5o W
; 1. Go to Figure 2.1 Enter the date on the form.
1
- e s C/
ODCM 6 Rev. 4
g-- 2. Enter the concentration C, (pCi/ml) for each isotope i, in a typical R LACBWR waste batch.
- 3. The values of f and F are determined and recorded at the top of l 3 e
i g Figure 2.1. F is the minimum volumetric dilution flow rate during releases ; i Ei
- g at the LACBWR - G-3 outfall which is equal to the LACBWR condenser o
( cooling water flow rate plus the G-3 condenser cooling water volumetric
.h ,
\ flow, in gallons per minute. The value f is the maximum radioactive liquid
.9 release flow rate (GPM) for the batches discharged during the period. A o
value of s17 GPM is normally specified for f. 8 k N
- 4. The quantities E C,, and E C,/MPC, are determined and recorded.
~~
W D The monitor conversion factor, k, determined at last primary calibration is y 5. ;
.e .
N ) y q recorded on Figure 2.1, in cps (net) per pCi/ml. 8 . l is l f d 6. The alarm setpoint, a (cps), with a 0.5 factor for conservatism, for the E m6nitor measuring radioactivity in the liquid effluent line is then determined by e E 8 0 " " 0.5 kF ECi i f ECi MPCi/ s i
? z
.e o
? $ b) To demonstrate compliance with the limitations of Section 3.2.2.b, dose 5 E l I
$ _i contributions are calculated at a maximum interval of once every calendar quarter for
$j all radionuclides identified in liquid effluents released to unrestricted areas using the A
U__ , 1 ODCM 7 Rev.4
~ .. - - . . -. .. - - - - - -- - .
i l
)
j __ methodology presented in NRC Regulatory Guide 1.109 Rev.1, October 1977. This ( o
. methodology takes the form of the following general equation:
r- D., = I(A,8 $ Cy/Fj) (2.4) g i 1-1
- a 1
9 < where: E i i 0 b D., = the cumulative dose commitment to the total body or any organ l 4 l ha s I of an individual in age group a from the liquid effluents released j 8 N in m batches, in mrem. gb E o N' ' Cy= the total quantity of radionuclide i, released by batch J, in Ci. t
, A g A,a = the site-related ingestion dose commitment factor to the total body 1
or any organ T of an individual in age group a for each identified s principal gamma and/or beta emitter, in mrem-gal-min '-Ci-' . f g
- 'E =
x , .L3 F, = the average dilution water flow rate during batch release j in sL% l h i! gallons / minute, ii
- Equation 2.4 requires the use of a dose factor Aoi, for each nuclide, organ and individual in age group a which includes the factors which determine the ultimate dose
~
O g received such as pathway transfer factors (e.g., bioaccumulation factors), pathway i [ usage factors, ingestion dose factors and dilution factors. The following philosophy l I E z s @ and site-specific conditions determine the site-specific factors incorporated into the
$ El Z
j g liquid effluent dose calculation model: h
! 5 o
V ODCM 8 - Rev. 4
1 l l 4
- 1. Liauid Dose Pathways l R ge, N
Due to LACBWR's status as a fresh water site, there is no invertebrate O pathway. The drinking water pathway is not included, since the nearest l community which obtains its drinking water supply from the Mississippi River is fe r located at Davenport, Iowa, which is 195 miles downstream. The drinking water pathway represents < 0.01% of the dose to any organ. The irrigated foods g s E f pathway is not included since the river water is not used for irrigation in this area and the shoreline deposits pathway is insignificant for the Mississippi river. The
,g '
only significant dose pathway is the dose commitment due to ingestion of fish
& 4 from the Mississippi River waters.
E \. fl, 2. Dilution U The liquid effluent flow from the waste tanks is diluted by the combined q total circulating water flow for condenser cooling at both LACBWR and G-3. For 5 $ offsite dose calculations, no dilution by the Mississippi River flow is considered. h Also, under SAFSTOR conditions batch discharges of liquid effluent normally take place during less than 35 hours per month (< 5% of the time). Therefore, 3 8 M m no fish in the river are continuously exposed to a radioactive environment yp produced by LACBWR liquid effluent as assumed in the calculation of the
.Qg published bioaccumulation factors for fish.
p c g kx i Based on the above site-specific criteria, the dose factor A,., is defined as e e E N follows: A,3 = K, (UF,)(BF,)(DF,3 ) (2.5)
$ $ l c e where-l 5 K, = a units conversion constant,5.03 E5 =
? z
$ O E $ (1 E12 pCi/Ci x .2642 gall t) / (8760 hrs /yr x 60 min /hr).
IE v i j g UF, = fish consumption usage factor for an individual in age group a, in em kglyr. E $ V ODCM 9 Rev.4
n BF, = the bioaccumulation factor in fish for nuclide i, in pCi/kg per pCi/t
'V y y4 D F,3 = the ingestion dose factor for age group a for nuclide i, in organ i in
& mrem /pCi.
g-
/
e
& Calculation of Dose Commitments from Liauid Effluents 9
E j , The equations for this calculation have been formatted on a LOTUS h spreadsheet. The values of UF,, BF,, and DF, specified in NRC Regulatory Guide a s j g 1.109 Rev.1, October 1977, and the constant K, have been entered on the spreadsheet. o
==
To perform the calculation the following information is entered in the appropriate h cells of the spreadsheet for each liquid batch released during the period of interest:
~~
Oj s 1. Date
$ 1 2. Release interval, hrs 8
gg 3. Waste volume, gal z
- 4. Condenser cooling water flow rate, GPM g
j 5. Activity concentration of each isotope, i, in waste, pCi/ml. The spreadsheet program will then calculate and display the total quarterly dose
]as O
w
$j in mrem to the total body and each organ of an individual in each age group. The
}~E z cumulative calendar year doses and the percentage of the limits set forth in j Section 3.2.2.b are also calculated. This spreadsheet will also print the data tables for a
j j the liquid effluent section of the annual report.
! 5 O --
ODCM 10 Rev. 4
Figure 2.1 C- < LIQUID RELEASE MONITOR j, g - ALARM SETPOINT DETERMINATION __ Date E ' [ Maximum Liquid Release Rate for Period, f= GPM
& Minimum Dilution Flow Rate for Period, F= GPM s
g 5 Average Concentra- MPCi(10 CFR; g , Nuclide 1 - tion (inTanks),;.. Part 20, Ap~pendix.B. J C,/MPC, g i C,(pCi/ml)L Table 2, Col.12)4
$ Co-60 3 E-06 5 AJ Cs-137 1 E-06 Ed -
gg Mn-54 3E-05 2- Ce-144 3E-06 4 Zn-65 5E-06 e o y Cs-134 9 E-07 Ru(Rh)-106 3E-06 1 Sr-90 5 E-07 I s 8 q Fe-55 1 E-04
> i
& Ag-110m 6 E-06 1
$ ECi = IC;/MPCi = l
; , 4 Q. h) b k*
- y. .
}
Monitor Conversion Factor, k = a o e a s; 0.5 W E C' = cps above background g (awrm setpoire) f IC /MPCi i
? z 2 8 e a b Z .
u a e s O ODCM 11 Rev.4
2.2 Compliance with the Limitations for Gaseous Effluent Releases l [ N a) To assure compliance with the limitations of Section 3.3.3.a, alarm setpoints are O
. established for the gaseous effluent monitor. These setpoints are calculated or m
checked annually, or as required by procedure, to confirm that the current setpoints are k g j set correctly for one- or two-stack blower operation. 5 o< During SAFSTOR, the offgas treatment system from the condenser to the stack
.f a E \
g is no longer in operation since the plant is shut down. The principal potential gaseous E 6 o {g re! case pathway is from the Containment Building ventilation exhaust system. The only 1 noble gas potentially available for release from the facility is Kr-85. The irradiated fuel 4 S assemblies stored in the Fuel Element Storage Well (FESW) contain essentially all o q% , the Kr-85 inv6 story. There is a very small potential for a Kr-85 release from the Waste , e_ E s
) Treatment and Turbine Building ventilation exhaust systems. This would be possible R
only if FESW water containing Kr-85 were transferred to the Spent Resin Receiving R g Tank (SRRT) or the Waste Water Tanks (WWT). Activity in particulate form and H-3 E { j s can theoretically by released via any of these release pathways. There will be no radiciodine (I-131,1-133) releases since they are no longer being produced and, since I shutdown, any residual activity has decayed to insignificant levels. H E 1 8 e s
$ E v 3 r!5 V 4 ODCM 12 Rev. 4 s
4 l l ,
, { r_ Noble cases -
i . I $ The following mathematical relationships shall be used to implement the above 4 , ! g requirements for noble gas (Kr-85) release alarm setpoints: E . a J k *I'
- ~ g D = K Q (x/Q)(DFB) i E
= K' Q,F, (x/Q) (DFB) (2.6)
! 3 .s Y D = K [1.11 Q(x/Q) DF + Q(x/Q)(DFS)] E l f{ = K' Q,F, (x/Q)(1.11 DFY + DFS) (2.7) where: 'n g S. A . 1 4 0 t D = the dose rate in mrem /yr to the total body of an individual beyond the 3 EFFLUENT RELEASE BOUNDARY due to Kr-85. This value is to be less j g g s than 500 mrem /yr.
'g ( K' = unit conversion constant,10 5pCi/pCi.
zc i ?
! , F, = volume flow rate in stack, cc/sec. ' + a.
- s Q= average Kr-85 release rate, pCi/sec.
! 2 I Q, = average Kr-85 release concentration, pCi/cc. i o (x/Q) = atmospheric dispersion coefficient for instantaneous releases. (For the FAST alarm setpoint,6.05 E-5 sec/m is used, based upon Regulatory S Guide 1.3 criteria. For the SLOW alarm setpoint,3.90 E-6 sec/m is used, based upon actual historical monthly average x/O values at the di worst case receptor location.) E z j $ DFB = the total body gamma dose factor for exposure to a semi-infinite cloud of l Kr-85 = 1.61 E-5 mrem-m per pCi-yr. jg l u J b* = the dose rate to the skin of an individual at or beyond the EFFLUENT j kh y3 RELEASE BOUNDARY due to Kr-85. This value is to be less than 3000 !
-- mrem /yr.
ODCM 13 Rev. 4 l
1 4 1.11 = the ratio of the tissue to air energy absorption coefficients over the energy {
]g y range of photons of interest. This converts dose (mrad) to dose j jg .
equivalent (mrem). ~
--- DFY= the gamma air dose factor for exposure to semi-infinite cloud of Kr-85 =
3 9 1.72 E-5 mrad-m' per pCi-yr.
- e
! E DFS = the skin beta dose factor for exposure to a semi-infinite cloud of j l Kr-85 = 1.34 E-3 mrem-m per pCi-yr. I \ NOTE: Equations 2.6 and 2.7 incorporate the use of the semi-infinite plume s; model. The model assumes receptor submersion in a plume of uniform E concentration, which is semi-infinite in geometry, having as its only y boundary the ground plane. Due to the meteorology and topography at
- R r the La Crosse Site, the worst receptor locations are a bluff 1300 m SSE O&( and a bluff 600 m ENE of the facility. At these locations the receptor is submerged in the plume.
! 4 l %' Calculation of Instantaneous Release Rate Monitor Setpoints for Noble Gases (Kr-85) O~g [ Equations 2.6 and 2.7 are used to calculate the controlling instantaneous N
- K E
i release rate setpoints for dose rates to the total body and skin of an individual due to i 8 e Kr-85 for one- and two-blower operation. i [
- E I
E g The DFB, DFYand DFS values for Kr-85 are multiplied by the appropriate X/O ! value, the conversion constants and the stack flow rate for one- or two-stack blower j
$ $ operation to obtain the values for TBF, SFG and SFB which are then inserted into the U ! N following equations to determine gaseous release monitor alarm setpoints, Q,:
! AI 4 e Q, (Whole Body) = I (2.8) TBF is Z ! u J (3000 mremlyr) i
= ( )
O. (Skin) (SFG + SFB)
& 5 !
J 0 ODCM 14 Rev. 4 l a
d _ l l i n V where:
- . w 3
3% TBF = (1E6) (x/Q) (DFB) (F,) SFG = (1.11E6)(x/Q)(DFY )(ps ) 7-I k j : SFB = (1E6) (x/Q) (DFS) (F, ) E
@ The smaller of the two values calculated is used for the setpoint. This instantaneous o .f release rate setpoint is very conservative since it is the average release rate allowed E
gg for a whole year. 8 & o i The following step-by-step procedure may be used in conjunction with the
==
, g equations found on Figure 2.2 to calculate the instantaneous release rate limits for 16 N.
$. Kr-85 for one- or two-stack blower operation:
s 1. On Figure 2.2, enter the date that the alarm setpoint calculation is 5 g performed. ;
.0 4 2. Note the appropriate value for x/Q.
st
- m i 3. Note the appropriate value of F, and the number of stack blowers :
j operating for the condition being calculated. l r !
- 4. Using the equations at the top of Figure 2.2, calculate the values of TBF.,
l SFG and SFB for one- and two-blower operation and for all appropriate I X/Q's, using the DFB, DFY and DFS values for Kr-85 listed on Figure 2.2. j $ a is _2 5. Calculate the values of Q, for each case. Select the smallest Q, value g for each x/O which will become the FAST and SLOW alarm setpoints for B z the noble gas monitor. a o B $ 6. The Q, values (alarm setpoints) for 1989 are tabulated on Figure 2.3. x w g z 3 $, NOTE: These values will not change during SAFSTOR unless the limits are EE changed, the X/O values are changed, the dose factors are changed or iE $ the volume flow rate in the stacks changes. ODCM 15 Rev. 4
H-3 and Particulates ee, o .' The following mathematical relationship shall be used to implement the limitation i- ' for H-3 and Particulates with Tm > 8 days alarm setpoints: e a . 5 De, = Ei Ph Qpi(x/Q) (2.10)
'E 5'
0 where: j ,% be, = the dose rate to organ T of an individual at or beyond the EFFLUENT RELEASE BOUNDARY, due to H-3 and particulates i 8@) with half-lives greater than 8 days. This value is to be less than b , 1500 mrem /yr. M Pu = the dose parameter for organ T , for radionuclide i, for the { inhalation pathway, in mrem-m per pCi-yr. o $( x/O = the atmosphere dispersion coefficient in sec/m
\ release rate of nuclide i, in pCi/sec. & O,=
g I t a h% [ Calculation of Release Limits for H-3 and Particulates s with Half-Lives Greater than 8 days 5 r + Since it is impractical to measure instantaneous release rates for radionuclides s @ other than noble gases, the alarm setpoints for radionuclides other than noble gases o 0 " are expressed in terms of total accumulated activity on sample media for a specified sampling time, AT, which is monitored as Ci by the stack effluent monitor. z
$ 8 e a g z ,
v a ee O ic 5-ODCM 16 Rev. 4
l l Equation 2.11 is used to calculate the release rate limit for all H-3 and Q particulates with half-lives greater than 8 days. This equation is based on the dose o rate to an infant due to inhalation of these radionuclides. In accordance with o e ;' NUREG-0133, the infant will always receive the maximum dose rate. The atmospheric E I
$ ss dispersion coefficients (X/Q) used are 6.05 E-5 sec/m for the calculation of the FAST E
f g alarm setpoint and 3.9 E4 sec/m for the SLOW alarm setpoint. R Alarm Setpoint Calculations for H-3 and Particulates g e with Half-Lives Greater than 8 days e o
% 1500 mreru/yr R (2.11)
Q O"' = I [Pa (inhalation) x R ](x/Q) pi o (c v (" s where:
'o g 5,
& Og,= the maximum allowed total release rate of a typical mixture of
.!3 N.
> radionuclides in pCi/sec conservatively derived from the allowed annual g y average dose rate to organ T and very conservative X/O.
8
- r. s Rg; = the ratio of the activity of nuclide i, to the total activity of all nuclides other g $ than noble gases in a typical mixture being released.
R X/ O = the atmospheric dispersion coefficient as given above for FAST or SLOW m la z alarm respectively, in sec/m'
$ 8 e d g z ,
u a ex
& 5 ,y O
ODCM 17 Rev. 4
Resolution of the P,, term in Equation 2.11 yields: Gs g. lii P,, (inhalation) = (10' pCi/pCi)(BR)(DFA,) i (2.12) where:
} DFA,, = the inhalation dose factor for an infant, for the i* radionuclide, for
{ organ t, in mrem /pCi. d s BR = infant breathing rate, in m'/yr. 5 0
) To calculate the alarm setpoint in terms of total pCi deposited on filter or h , cartridge sample media, the following equation is used:
e 0 Lowest Op , x AF O" =
% F.
2 N g { where: O, = the activity in pCi (deposited on sample media in sample time AT) which g' initiates an appropriate alarm in the stack effluent monitor. 5
& Og, = pCi/sec 8%
F, = stack flow rate, cc/sec
$ AF = total flow through sample media (cc), in sample time AT, corrected to i s stack gas conditions. AT is normally 7 days.
i S E !
@ @ l E The procedure outlined below is used to calculate the release limits for j E radionuclides other than noble gases. This will be done at least annually.
b o s jE NOTE: This procedure is applicable for the determination of either FAST or 3 $ SLOW alarms by utilizing the appropriate value for x/O in the equation. l E$ a 5 ; O ; l ODCM 18 Rev. 4
- 1. Start on Figure 2.4 Enter the date, the alarm setpoint being calculated N FAST or SLOW) and the appropriate x/O value to be used.
is
.[ 2. Enter the average release rate for the period, On, in pCi/sec, of each identified radionuclide. At the bottom of the form, compute and enter the
_h a I sum IQpi . E E' 8 1
- 3. In the column labeled Rn, enter the ratio of the average period release l rate of nuclide i to the average total period release rate, IQpi , for the m , i E\ period.
p% g
> 4. For each organ T, as noted at the top of the form, calculate and enter the
% value of (x/Q) (Rn) P,,(inhalation) for each nuclide. Pa(inhalation) values 2 N
$ are found on Table 2.1. At the bottom of the column, for each organ,
{ enter the value of I Rei Pa (x/0) for that organ, (~' v j [ 5. Go to Figure 2.5. Enter the date and the alarm setpoint being determined. s { 6. Using the equation at the top of Figure 2.5, calculate the release rate i s limits, Og, , for each organ T. 3 $ 7. Select the lowest value of Og,, enter at the bottom of Figure 2.5 under 5 0 _ E appropriate blower operation. Multiply the Or , number times the total z sample flow through the sample media, cc, and divide this by the
.e o h$ appropriate blower flow rate, cc/sec, to determine the Q, in pCi and use ti z ,
j d these as alarm setpoints.
&5 m__
L) ODCM 19 Rev. 4
l l l I
\
! b) To demonstrate compliance with the limitations of Section 3.3.3.b, dose g__a I g contributions are calculated for any Kr-85 released to unrestricted areas using the l O s following expressions: E [ , f DY (r,0) = 3.17 E-2 DFY Q /Q] M 0) (2.14) l ' Da (r,0) = 3.17 E-2 DF" Q [x/Q](r,0) (2.15) 0 g where: 5m DY(r,0) = the dose commitment to the maximum individual due to the g gamma radiation from Kr-85 at location (r,0), in mrad. g DS (r,0) = the dose commitment to the maximum individual due to the beta radiation from Kr-85 at location (r,0), in mrad. 9 e R ' Q = the total release of Kr-85 in gaseous effluents for the release 8R period, in pCi. j i 3.17 E-2 = pCi/pCi divided by sec/yr p -- v g-s t [X/ Q](r,0) = the annual average atmospheric dispersion constant for long-
@ term releases at location (r,0), in sec/m . Since the collection
.g g of hourly meteorological data is no longer required or
; performed at the LACBWR site, a conservative value based on i historical site specific annual average x/O values will be used.
j This value is 1.82E-6 sec/m'. 1, s DFYand DF" = the gamma and beta air dose factors for exposure to a uniform semi-infinite cloud of Kr-85 in (mrad-m'/pCi-yr). Numerical g o g l values are 1.72E-5 and 1.95E-3 respectively. (Ref. NRC Regulatory Guide 1.109 Rev.1, October 1977) 3 5 e s
$ E E
5h
!5 0 --
ODCM 20 Rev. 4
Calculation of Gamma and Beta Air Dose Commitments a h w in accordance with the RECP, the gamma and beta air dose commitments are to be calculated once per calendar quarter and yearly. Equations 2.14 and 2.15 are used
}
e f to perform these calculations. Since the only noble gas that needs to be considered at a 9 LACBWR is Kr-85, and since a conservative constant value is used forx /Q, these E E g equations reduce to: E h . D7 = 9.923E-130 E eT DD = 1.125E-100 EA3 h% g ox ( The following step-by-step procedure is used in conjunction with Figure 2.6 to D calculate the quarterly cumulative dose commitments due to Kr-85. O v g N
- 1. Go to Figure 2.6. Enter the Date. Enter the period covered by the 3
g 4
. calculations.
i i E 2. Enter the total Kr-85 activity released in the gaseous effluent during the
$ s period being considered, in pCi.
e
$ $ 3. Calculate the dose commitments DY(r,0) and D"(r,0) due to Kr-85 using o e the equations on Figure 2.6.
E IB z
.g o 4. Calculate the percent of the current quarterly and annual technical E id b Z specification limits and enter on Figure 2.6.
v > c 5 O O ODCM 21 Rev. 4
q g _._ c) To demonstrate compliance with the limitations of Section 3.3.3.c, dose ( c'
. contributions are calculated for H-3, and particulates with half-lives greater than 8 days,
- identified in gaseous effluents released to unrestricted areas using the methodology 7 '
$ presented in NRC Regulatory Guide 1.109, Rev 1, October 1977. This methodology e
takes the form of the following general equation:
}.
5 0 D,,(r,0) = E E Mfs, W(r,0) Oi (2.16) g Pi
.h3 where:
E , b D,,(r,0) =the dose commitment to organ T of an individual in age group a, at distance r in sector 0 from the release point, due to the release to the y atmosphere of radionuclides other than noble gases, in mrem. S o W(r,0)= the average dispersion parameter for estimating the dose to an (* individual at the receptor location (r,0), for the period of release, in g g s sec/m or m-2 as required by the characteristics of the exposure pathway.
?!
4 = the total activity of each radionuclide i, other than noble gases, in Q, gaseous effluents for the release period of interest, in pCi.
.( ,
s = the dose conversion factor for exposure pathway P to organ i of an e Mft, i individual in age group a, for each identified radionuclide i. The units I .
- of Mf, are (mrem-m 2)/pCi or (mrem-m )/pCi-sec) as required so that the product Mf,, W(r,0) is mrem /pCi.
2 E o 0 n g Equation 2.16 may be expanded to the following form where each term is the E z
.g o incremental dose received via one of the three major dose pathways.
$- Ei z
@ .) D,, (r, 0) = I DR, (r,0) + D$ (r, 0) + D , (r, 0) (2.17) a $
C) U ODCM 22 Rev. 4
where the first term on the right is the external dose from direct exposure to activity g __. i l N U deposited on the ground plane, the second term is the dose frr .n inhalation of radionuclides in air, and the third term is the dose from ingestion of foods contaminated J a
$ by atmospheric releases of radionuclides.
e l 9 E g Applying the methodology of NRC Regulatory Guide 1.109 Rev.1, equation 2.17 0 1 "\ is expanded as fo!!ows:
?
a:
,j ( D,, (r, 0) = E MR, Qi(D/Q)(r,0) (2.18) fU2
+ E Mf Oi(x/Q)(r,0) i e
d x: 5 + E Mf,V Qi(D/Q)(r,0) + (M?L Qi4 + M?V, OT ) (X/Q)(r,0) D i
+ E MS,* Qi(D/Q)(r,0) + (MP4 Qi4 + M?" QT)(X/Q)(r,0)
.f ~ '
5 l
.0 A '
I + E MR7 Qi(D/Q)(r,0) + (MfL 014 + id?f, Or ) (x/Q)(rf,
- n. i ,
E 8 I -
* + E Mfj Qi(D/Q)(r,0) + (M?h Qi4 + M?), O1) (x/Q)(r,0) l 2 I@
5 $ where: g (x/Q)(r, 0) = the annual average atmospheric dispersion factor for a receptor at 1B z the distance r in sector 0 from the release point, in sec/m' For y 8 the LACBWR in the SAFSTOR mode, the value for this term is
& Ei conservatively taken to be the largest historical (1963-1987) o Z . undecayedlundepleted x/Q for a real receptor and is 1.82E-6 3 d sec/m .
- aE l r 5 l
O ODCM 23 Rev.4
(D/Q)(r, 0) = 1.82E-9m.2 This is based on the relationship D/O = V, x/Q h-g where V, = the deposition velocity in m/sec. V, is generally
< 1 E-3m/sec for dry deposition of submicron aerosols which may g be released from the LACBWR facility during SAFSTOR (Ref Whicker, F. W. and Schultz, V., Radioscology: Nuclear Energy 1 and the Environmerit, Vol 11, CRC Press, Inc., Boco Raton, 5 Florida,1982.
E \'
% M,i = 1.0E6 Sr DFG3, (1-e-%)/Ai and according to R.G.1.109 the s '
i dose to all internal organs (t) for all age groups (a) is taken to be f i the same as the total body dose. J A M, = 3.17E-2 BR,DFA,, a: E ik and for the ingestion pathway (DV) for produce (non-leafy-vegetables, fruits, and grains) M i.V = 1.1E2 DFli UY fg exp(-Ait n)(r(1 -exp(-Atit.))/ YvAti + o g ( Bw (1-exp(-Aito))/ Pli) E'd for all radionuclides except C-14 and H-3 Og N M f. = 22 DFly , U;fg p for C-14 3 t 5 MEX = 12 DFir U;f,/H for tritium
-f f
g for the ingestion pathway (Dm) for milk 8 E
- Mi f = 1.1E2 DFI,, UT Fmi Oe exp(-Ai tr)(fp f. (1 - exp(-li tn }) + exp(-Aiin ))
-- x (r(1-exp(-Acit.))/YvAti + Biv(1 -exp(-lite))/PAi) ;
e 8 for all radicruclide except C-14 and H-3 ! j @ n I- MD4*, = 22DFl , yUT Fmi Or p(exp(-Ai4tr)) for C-14 m E o M?*. = 12DFl , yU* Fmi Opexp(-AT tr)/H for tritium E $
$ E y i a
2 =5 O ODCM 24 Rev.4
for the ingestion pathway (DM) for meat G- , M,P," = 1.1E2 DFli. UY Fn Orexp(-lit.) (fpf.(1 -exp(-litn)) + exp(-Aith) ) x (r(1-exp(-Ati t.))/Yv Asi+Biv(1 - exp(-Aitb ))/Pli) for all radionuclides except C-14 and H-3 __s M M = 22 DFl 4.UP Fr.4 Or p(exp(-134t.)) for C-14 k 4s. i I M?u. = 12 DFir UP FrrQFOxp(-ATts)/H for tritium 8' Tr for the ingestion pathway (DL) for leafy vegetables: e
. - Ml,i= 1.lE2DFIiuUIf,exp(-lith)(i(1 - exp(-Ar;t.))/YvAr,+Biy(1 - exp(-A,ts))NAi )
for all radionuclides except C-14 and H-3. o e 6 g ME}, = 22DFlumUkf,p for C-14 es
)) Myh = 12DFIroUkf,/H for tritium J y ] q The values used for the various parameters in the above equations are those gN recommended in NRC Regulatory Guide 1.109, Rev.1, for the maximum exposed E
g individual. 8 I g Parameter Dimensions Description / Source e
$ $ 1.0E6 pCi/pCi E
I- DFGa mrem-m2 /pCi-hr from table E-6 in R.G. . S ) E z l s8 DI A mrem /pCiinhaled from table E-7 thru E-10 in R.G. 5- Ei z 1 _; DFAuw mrem /pCi inhaled from table E-7 thru E-10 in R.G. e s o__ L) ODCM 25 Rev. 4
l y __ Parameter Dimensions Description / Source { N
- c DFA rw mrem /pCiinhaled from table E-7 thru E-10 in R.G.
E DFI,w mrem /pCiingested from tables E-11 thru E-14 in R.G. 5 4 ( DFli4w mrem /pCiingested from tables E-11 thru E-14 in R.G. l g DFIrw mrem /pCi ingested from tables E-11 thru E-14 in R.G. l 0 )
.$ N Sr = 0.7 dimensionless attenuation factor accounting for i
[ Nq shielding by residential structures I E ' e L hr-! radiological decay constant for O [ nuclide i. 1 5 to = 1.31x10 hr period of long-term buildup for
, 2 activity in soil (nominally 15 yrs)
EN 3.17x10-2 pCi-yr/pCi-sec 9 g s B R, m3 /yr inhalation rate for age group a. i t Table E-5 in R.G. e
% ,. 1.1x102 pCi-pr/pCi-hr R
h Uy kg/yr consumption rate of produce for 8 individual in age group a. Table E-5 I ' of R.G.
, g f, = 0.76 dimensionless fraction of produce ingested that is g g grown in garden of interest.
n t, hr he &by MMe Mmd d E vegetation or crops and ingestion. 1B z
$ 8 =0 for pasture grass by animals 5- Ei z
Ei = 2160 for stored feed by animals E E5 (m G ODCM 26 Rev. 4
r-U Parameter Dimensions Description / Source { : gy = 24 for leafy vegetables by man a N - __ = 1440 for produce by man ! l i k r = 0.2 dimensionless fraction of deposited activity ! retained on crops, leafy vegetables, ) E or pasture grass. I 0 g g As, = A, + Am hr' the effective removal rate constant X for radionuclide i from crops. e m
$ l Ar =.0021 hr' removal rate constant for activity on plant or leaf surfaces by weathering o k (~ to 14 day half-life) g t, hr period of crop, leafy vegetable, or S
a b pasture grass exposure during growing season.
~
= 720 for grass-cow-milk-man pathway
.e j % = 1440 for crop / vegetation-man pathway 8
} ' Y, kg/m 2 agricultural productivity (measured in wet weight) i is
$ 4 = 0.7 for grass-cow-milk-man pathway e = 2.0 for produce or leafy vegetables j $ ingested by man U
N
, B, dimensionless pCi/kg in vegetation per pCi/kg in m soil for nuclide i. Table E-1 in R.G.
E z
- o
.j y P = 240 kg/m 2 effective surface density of soil (dry 3 z ,
weight) e a e >-
}@ 22 pCi -yr - m'l Ci- kg - sec
& 5 O
ODCM 27 Rev. 4
Parameter Dimensions Description / Source { L_ N c p dimensionless the ratio of the total annual release 5 time for C-14 to the total annual time during which photosynthesis occurs y with the condition that p51.0 e e 4 .
= 1.0 for continuous C-14 releases.
E 1 E s ~ 0 12 pCi-g-yri Ci-kg-sec H = 8.0 g/m' average absoluts humidity of the ; [ g , atmosphere at location (r,0) l 0& consumption rate of milk for { o N UT liters /yr individual in age group a. Table E-5 of R.G. g Fu day /l factor for estimation of activity of C Q nuclide i in milk from that in ariimal (> feed (pCi/ a milk per pCi/d ingested by the animal) Table E-1 in R.G. j s Og = 50 kg/ day feed or forage consumption rate !
/d (wet weignt) by milk cow or beef cattle .g ,
i s f t, = 48 hr transport time from animal {c feed-milk-man. l 1
* * = 0.5 dimensionless fraction of the year that animals i f,
graze on pasture. e y $ f, = 1.0 dimensionless fraction of daily feed that is pasture o M when the animal is on pasture E UP kglyr consumption rate of meat & poultry for individual in age group a. Table
] g jg E-5 of R.G.
EE
$i es O
ODCM 28 Rev.4
g__ Parameter Dimensions Description / Source o Q Fn day /kg factor for estimation of activity of , o nuclide iin meat from that in animal ! QU feed (pCi/kg in meat per pCi/ day 3 ingested by the animal) Table E-1 in :
$ / R.G.
e j , 9 t, = 480 hr average time from slaughter of meat animal to consumption of meat 0 1 g o U[ kglyr consumption rate of leafy y vegetables for individualin age m group a. Table E-5 in R.G. l
. f, dimensionless fraction of leafy vegetables grown in
= 1'0 garden of interest.
O gg , e A ax o( Calculations of Dose Commitments due to Gaseous Release other than Noble Gases L/ y s in accordance with the RECP, the maximum commitment to a MEMBER OF THE l
.e PUBLIC from H-3 and all radionuclides in particulate form with half-lives greater than 8 i m
E ^ 8 days shall be determined at least quarterly.
- o. ,
c
%o I % To perform this calculation Eq 2.18 has been formatted on a LOTUS
, spreadsheet. The quantity in curies of each nuclide (i) released to the atmosphere 4 Q from the LACBWR facility during the calendar quarter is entered in the appropriate cell R
[z of the spreadsheet. The spreadsheet program calculates and displays the total 7 j } quarterly dose in mrem to the total body and each organ of an individual in each of four E z f, ; age groups and the cumulative calendar year dose to the total body and each organ. It
@m also determines the maximum exposed organ (and its dose) for each aqe group each
- a. 5 LJ ODCM 29 Rev. 4
i ( j q __ quarter and the dose to the maximum exposed organ in all age groups. The quarterly g ! and cumulative calendar year doses to the maximum exposed organ are compared to f a O ! the limits and the relation in terms of percent of the limit is displayed. The maximum i _. o
> incremental organ dose received through each of the three major pathways is also i . determined for each age group each quarter.
E Y J $ o ' I E.. ' i '. E
. .9 .
E k , 4
, s i ab i oq
- V
- m --
!. b .g , E ' i- o:
'k%
!- f f i e i 3
, I
- l ==
. 8
- a m ,
- O e n
i __
- m i l 3 $ I s e u E d -
{ g z , l
- .a l l
1 >- I E j ! @$ i i O ODCM 30 Rev. 4 i l i i J
- + , ,- , ,..--g - ,-n , ,r-
O Prepared or Revised By Date O Health Physics Revi_ew Data Operations 13; view Comm. Approval Data . el LARRY L. NELSON 3/25/96 f-( 8/ /// /9(, // C[ /j M h 9/#/0Yl Table 2.1 INFANT DOSE FACTORS Pa(INHALATION) FOR H-3 AND PARTICULATE GASEOUS RELEASE MONITOR ALARM SETPOINT DETERMINATIONS in Units of mrem-m2 /pCi-yr Nuclide Whole Body Bone Liver Thyroid Kidney _ Lung GI-LLI H-3 6.47 E2 6.47 E2 6.47 E2 6.47 E2 6.47 E2 6.47 E2 MN-54 4.98 E3
- 2.53E4
- 4.98 E3 1.00 E6 7.06 E3 C0-60 1.18 E4 8.02 E3 4.51 E6 3.19 E4 ,
ZN-65 3.11 E4 1.93 E4 6.26 E4 3.25 E4 6.47 ES 5.14 E4 SR-90 2.59 E6 4.09 E7 *
- 1.12 E7 1.31 E5 RU-106 1.09 E4 8.68 E4 *
- 1.07 E5 1.16 E7 1.64 E5 CS-134 7.45 E4 3.96 E5 7.03 E5 1.90 ES 7.97E 4 1.33 E3 CS-137 4.55 E4 5.49 ES 6.12 E5
- 1.72 ES 7.13 E4 1.33 E3 9.84 E6 1.48 E5 CE-144 1.76 ES 3.19 E6 1.21 E6 5.38 ES Values in this table are derived from Tables E-5 and E-10 in App. E of NRC Regulatory Guide 1.109 Rev.1, October 1977.
- No data available.
ODCM 31 Rev. 4
i l Figure 2.2 I b\ l yh NOBLE GAS (KR-85) RELEASE MONITOR ALARM l SETPOINT CALCULATIONS l
} Calculation for Alarm Condition (FAST or SLOW) l E x/Q = No. of Stack Blowers = F, = l j
d/ Ecuations: i TBF = (1E6)(x/Q)(DFB)(F,) ) s
& Y a: SFG = (1.11 E6)(x/Q)(DF )(F,)
E 4 f&%h SFB = (1E6) (x/Q) (DFS) (F,) o where: o e R (X/ Q) = 6.05 E-5 sec/m or 3.90 E-6 sec/m2 5 (y (FAST) (SLOW)
*~*'
e-~ } g DFB = 1.61 E-5 "'*Ci-yr p 3 s E DFY = 1.72 E-5 mrag-m'
.g 4 pCi-yr 2
I DFS = 1.34 E-3 "f**~*' g - pCi-yr
$ r F, = 1.65 E7 cc/sec or 3.304 E7 cc/sec (1 blower) (2 blowers) e 8 mrem- c 3 $ TBF =
n pCi-yr 5 mrem-cc SFG+SFB = + = z pCi/yr
> 0 Q,(whole body) = = pCi/cc
'*F e >-
- Y=
b$ Q"(skin) = pCi/cc E$ (SFG + SFB) ODCM 32 Rev. 4
i i l h . Figure 2.3 1 E / NOBLE GAS (Kr-85) RELEASE MONITOR ~ g ALARM SETPOINT
SUMMARY
- --- (pCi/cc in stack effluent)
# i e - !
i d 4 g FAST ALARM SETPOINT SLOW ALARM SETPOINT o O~ l 1 BLOWER 2 BLOWERS 1 BLOWER 2 BLOWERS h a: - E WHOLE BODY 3.11 E-2 1.55 E-2 4.82 E-1 2.40 E-1 S D e jk SKIN 2.21 E-3 1.10 E-3* 3.43 E-2 1.71 E-2* A
$ b.
- Since Kr-85's beta dose equivalent component is significantly higher than its gamma S '
dose component, the noble gas monitor alarm setpoints will always be based upon p D g-the Q, for skin dose. Since the alarm setpoints for 2 blowers are the most ! '5 i l ja % restrictive, they may be used for all operating conditions without exceeding the limits !' [ for instantaneous release. c ' is i . u 4 e 8
% D5 R
t
- 3 5 t S
$ Ei v
! 5 O --
ODCM 33 Rev.4
O Prepared or Revised By Data H:alth Physics R3 view e Data Operations Ppview Comm. Approvil Data T LARRY L. NELSON 312 5196 m[ if/II/7(1 /{(' [fuyK G/p,&{ l Figure 2.4 Date H-3 AND PARTICULATE GASEQUS RELEASE MONITOR ALARM SETPOINT DETERMINATION Alarm oeing calculated (FAST or SLOW) x/Q sec/m' * (x/Q) RriPis(inhalation) ** Nuclide i On R,p W Body Bone Liver Thyroid Kidney Lung GI-LLI H-3 Mn-54 Co-60 Zn-65 Sr-90 Ru(Rh)10 Cs-134 Cs-137 Ce-144 EQri = E= i i
- For FAST alarm use 6.05 E-5 sec/m for X IQ and for SLOW Alarm use 3.90 E-6 sec/m
** P., (inhalation) values found in Table 2.1.
ODCM 34 Rev. 4
Figure 2.5 4 F-\ H-3 AND PARTICULATE GASEOUS RELEASE ' ~ g MONITOR ALARM SETPOINT
SUMMARY
Calculation for (FAST or SLOW) alarm. 7g , [ 1500 mrem /yr g* ~_ 2 Ei [Pa(inhalation) x Rei x x/Q] d j O n = maximum allowed total release rate, pCi/sec to meet dose rate limit to organ T.
- g x I(x/Q) Rei Pa(inhalation)* On (pCi/sec) hC
, jy Whole Body o Ek Bone l __ Liver 4 8 ( Thyroid Kidney (
> Lung GI-LLI
'Y
&
- From Figure 2.4
.g g Two-Blower Operation j One-Blower Operation l
' l j ~ Lowest Qp, x AF
_ Lowest Qp, x AF
~
" " 3.304E7
$ % 1.650E7 1 I l
== where: l l AF = corrected total flow through l 0, = uCi/sec x cc 3 $
8 $ sample media, cc l 3.304 E7 cc/sec I l x l
= Ci i z Ou = uCi/sec x cc l )
.g o 1.650 E7 cc/sec l (Stack Particulate Alarm Setpoint) e d I bz = pCi l i v a l e >- l
- Alarm Setpoint)l a @ (Stack Particulate .
&5 O
ODCM 35 Rev. 4
O Prepared or Revised By Date Health Physics Review O Date Operations RpviewAomm. Approv:1 Data e LARRY L. NELSON 3/25/96 /hG f- M M// t/////76 g{d' W _ M/##{ f / Figure 2.6 Date AIR DOSE COMMITMENT TO A MEMBER OF THE PUBLIC FROM NOBLE GAS (Kr-85) RELEASE , Release Period Total Kr-85 Act. Released, Q = pCi Dose Calculation for Release Period D7 = 9.923 E-13 O = mrad S D = 1.125 E-10 Q = mrad Limits
' Gamma Beta Particle Calendar Quarter 5 5 mrad 510 mrad Calendar Year 510 mrad 520 mrad Current Air Dose Commitment Record for Calendar Year
. ist Quarter 2nd Quarterl . 3rd Quarteri ' 4th ' Quarter'--
. Calendar Year -
Gamma Beta Gamma Beta Gamma Beta Gamma Beta Gamma Beta Dose, mrad
% of Limit ODCM 36 Rev. 4
h , 3.0 RADIOACTIVE EFFLUENT CONTROL PROGRAM Q .; __ 3.1 Proaram Requirements
- !i e
1 e The Radioactive Effluent Control Program (RECP) shall conform to the guidance i
! of 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses 0 '
j to MEMBERS OF THE PUBLIC from radioactive effluents as low as reasonably E N
$\
achievable. This program shall establish the requirements for monitoring, sampling ; l E h and analysis of radioactive gaseous and liquid effluents released from LACBWR to , h% ensure the concentrations in effluents released to areas beyond the EFFLUENT g f RELEASE BOUNDARY conform to 10 CFR Part 20, Appendix B, Table 2, Columns 1 l Q % F and 2. It shall provide limitations on the annual and quarterly dose commitment to a O V Q ' MEMBER OF THE PUBLIC from radioactive liquid effluents in conformance with g , s s E Appendix 1 of 10 CFR Part 50.
.{g i
s The limitations of operability of gaseous and liquid monitoring instrumentation, 5
% including surveillance test and setpoint determination in accordance with Section 2.0, e Offsite Dose Calculations, will be included in this program.
- W Requirements for the Containment Building Ventilation System, including "g Z filtration and elevated stack release of exhausted air is included in Section 3.3.1.
s e 5 Ei 8 z E Y BE a 5 O ODCM 37 Rev.4 i
i g__ in accordance with provisions of 40 CFR 190, the restrictions and surveillance g requirements for total dose to any MEMBER OF THE PUBLIC from all LACBWR O related sources and dose pathways are presented in Section 3.4. r-S E E d 3.2 Liauld Effluents - g o g 3.2.1 Samplina and Analysis J g x All liquid effluent releases at LACBWR will be in batch form. A batch gs release is the discharge of liquid wastes of a discrete volume. Prior to sampling o
== for analysis, each batch shall be isolated and then thoroughly mixed, to assure w
$ I R
representative sampling. The radioactive content of each batch of radioactive _I liquid waste to be discharged shall be determined prior to its release, as per the - Q ' R following table: i d x 5 ! i c iie ! I s
- E E
0 E 3 8 s Y N 1i k5 O ODCM 38 Rev. 4 l
I 7 Table 3.1 hk RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS REQUIREMENTS FOR BATCH RELEASES 5 Ea 9 .. . - .. .
. . - ._ :LSAMPLING- MINIMUM f:
l O (TYPE OF ACTIVITY' ANALYSIS A - { FREQUENCY) ANALYSIS FREQUENCY: O 1. Principal Gamma emitters N Prior to discharge Prior to discharge 1 2. Dissolved and entrained gases One batch / month Monthly E (gamma emitters) h 3. Tritium Prior to discharge Monthly Composite (') h 4. Gross Alpha Prior to discharge Prior to discharge Monthly Composite (*) Quarterly Composite (')
- 5. Sr-90 and Fe-55 Beta emitters 2
o (a) A composite sample is one made up of individual samples which are proportional to ( the quantity of liquid waste discharged and in which the method of sampling n employed results in a specimen which is representative of the liquid release. l V N (b) The principal gamma emitters for which the LLD specification will apply are exclusively
] t the following radionuclides: Mn-54, Co-60, Zn-65, Cs-134, Cs-137, and Ce-144. This E list does not mean that only these nuclides are to be detected and reported. Other
.h % s peaks which are measurable and identifiable, together with the above nuclides, shall y also be identified and reported.
c. Ei (c) Methods of calculating the Lower Limits of Detection (LLD) shall be contained in plant j procedures and are calculated in accordance with criteria of NUREG-0473, Rev. 2. o G
" The results of pre-release analyses shall be used in accordance with the Offsite Dose Calculations methodology to assure that the concentration at the point of z
n yd release is maintained within the limits specified in this RECP. 8 Z IB d aE l I
&5 O
ODCM 39 Rev.4
3.2.2 Liauid Effluent Release Limitation
\:
g4 a) Concentration - the concentration of radioactive material released in E % e liquid effluents at any time to areas beyond the EFFLUENT RELEASE
-m g BOUNDARY shall be limited to concentrations specified in 10 CFR e
4 Part 20, Appendix B, Table 2, Column 2, for radionuclides other than E. - E & O dissolved or entrained noble gases. For dissolved or entrained noble ! gases, the concentration shall be limited to 6E-4 ci/mi total activity
}
}( concentration.
l ib If the concentration of radioactive material released beyond the h% 9 EFFLUENT RELEASE BOUND /RY exceeds the above limits, restore the 2 R S k concentration to within the above limits without delay. t 8 g s This limit is provided to ensure that the concentration of radioactive Q materials released in liquid waste effluents from the site will be less than f E the concentration levels specified in 10 CFR Part 20, Appendix B, 2 Table 2, Column 2. The concentration limit for dissolved or entrained I g
- noble gases is based upon the assumption that Kr-85 is the controlling e
$ $ radioisotope and its MPC in air (submersion) was converted to an U
N equivalent concentration in water using the methods described in m
] g$
E International Commission on Radiological Protection (ICRP)
$ E Publication 2.
$ -i N5 g
\,_,/
ODCM 40 Rev. 4
q b) Dose - the dose or dose commitment to a MEMBER OF THE y PUBLIC from radioactive materials in liquid effluents released to areas 5. beyond the EFFLUENT RELEASE BOUNDARY shall be limited to:
=~
g . [ r Calendar Quarter
< 1.5 mrem total body Ca'endar Year 5 3 mrem total body E
e 5 5 mrem to any organ $ 10 mrem to any organ O 5 9 E 'N The cumulative dose contribution from liquid effluent shall be E j y\ determined at least once per calendar quarter in accordance with E O Section 2.0, Offsite Dose Calculations. If this calculated dose exceeds
" the above limits, prepare and submit to the Commission, within 30 days, a n
h Special Report which identifies the cause(s) for exceeding the limit (s) and s defines the corrective actions which have been or will be taken to assure g
'5 ; @ that subsequent releases shall be in compliance with the above limits. .QN ?
g This limit is provided to implement the requirements of Sections ll.A, ii Ill.A, IV.a and Annex of Appendix 1,10 CFR Part 50. The dose calculations in Section 2.0 implement the requirement in Section Ill.A of 2 E
$ y Appendix l that conformance with the guides of Appendix l be shown by g calculational procedures based on models and data, such that the actual 1B z y 0 exposure of an individual through appropriate pathways is unlikely to be 5 Ei 15 Z substantially underestimated.
v a
& 5 g
L) ODCM 41 Rev. 4
l P 1
. 3.2.3 Liauid Effluent instrumentation i b ,
l g The following radioactive liquid effluent monitoring instrumentation channels shall be OPERABLE, with their alarm setpoints set to ensure that the r~
- limits of Section 3.2.2.a are not exceeded, at all times when releasing liquid t
e [ } radioactive effluents. i 5 -t o i g + Liquid Radwaste Effluent Line Monitor or Turbine Condenser
- 5 p Cooling Water Monitor, and l E .
- e i
+ Liquid Radwaste Effluent Line Flow Meter.
j e
- o -'
l 1 The alarm setpoints for these monitors will be determined and adjusted using d o N' b methodology in Section 2.0, Offsite Dose Calculations. e
> l I - l lQ
- s The radioactive liquid effluent instrumentation is provided to monitor the
! .8, ' releases of radioactive materials in liquid effluents during actual or potential ; . e i i
-h E
releases of liquid effluents with the alarm setpoints set to ensure that the alarm
- g will occur prior to exceeding the limits of 10 CFR Part 20.
I 2 I ' a) Surveillance Requirements - each radioactive liquid effluent f I e 8 i w in monitoring instrumentation channel shall be demonstrated OPERABLE by [ R j , performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL 4 m , j $ FUNCTIONAL TEST, and CHANNEL CALIBRATION operations at the 8 $ y E frequencies shown in the following table. u a $ x
; k $
.b ODCM 42 Rev.4 l
h m
\. Table 3.2 gh RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION (D SURVEILLANCE REQUIREMENTS 9
E S ' Channel - d Instrument . Channel - L Source'- ' Functional L . Channel h . Check: LChecks ; Test 1 :Calibratio
.$ Liquid Radwaste Prior to Prior to Quarterly At least once Eg Effluent Line Monitor discharge discharge (See Note 1) per 18 months (See Note 4) (See Note 3)
$ 4 Turbine Condenser Prior to Monthly Quarterly At least once O
Cooling Water Monitor discharge (See Note 4) (See Note 1) per 18 months (See Note 3)
% Liquid rladwaste (See Note 2) N/A N/A At least once e Effluent Line Flow per 18 months t'
a Meter (See Note 5) Q v [v ) s NOTES: g-j t (1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room E alarm annunciation occurs if any of the following conditions exist:
- Instrument indicates measured levels at the alarm setpoint.
.h ,
+ Instrument indicates a downscale (circuit failure) failure.
{
$ (2) CHANNEL. CHECK shall consist of verifying indicttion of flow during periods of f g release. CHANNEL CHECK shall be made at least once per 24 hours on days in which continuous, periodic, or batch releases are made.
e (3) The CHANNEL CAllBRATION shallinclude the use of a known liquid radioactive
$ $ source positioned in a reproducible geometry with respect to the sensor. The Q source will have the gamma emitting radionuclide mixture and activity concentration which would normally be measured by the channel during batch discharges.
E (4) Background radiation may be used for the source check. E z M O E $ (5) The CHANNEL CAllBRATION will be in accordance with plant procedures. E z
$ i
$N O
ODCM 43 Rev. 4
/"N b) Corrective Action V g i N o (1) With the Liquid Radwaste Effluent Line Monitor or Turbine a Condenser Cooling Water Monitor channel alarm / trip-point setpoint e ; t less conservative than that required by Section 3.2.2(a), s Il e 1 immediately suspend the release or declare the channel inoperable o g j t or change the setpoint so that it is acceptably conservative. e (2) With both channels not OPERABLE, or if both alarm [CN o k setpoints are found to be less conservative than required, suspend 4 release of liquid radioactive effluent without delay. Effluent e$ 0 releases may be resumed with neither activity monitor OPERABLE, provided that at least two independent samples are analyzed and 4 I that at least two technically qualified members of the staff I s independently verify the release rate calculations. If channels are h f g not operable for more than 30 continuous days, explain in the next 8 I ' Annual Effluent Report. e
$ $ (3) With the flow meter not OPERABLE, effluent releases via M
this pathway may continue provided the flow rate is estimated at
] g$
B least once per 4 hours during actual releases. Pump curves may
$ j be used to estimate flow.
$ i N$
O ODCM 44 Rev. 4
4 . p 3.3 Gaseous Effluents ' 3& 3.3.1 Containment Buildina Ventilation J
- y Normal air discharge from LACBWR is made as an elevated stack e
$ release. Air is swept through the Turbine and Containment Building and then d ,Y E /,9 discharged out the stack. Whenever the Containment Ventilation dampers are o
open, the air from the Containment Building shall be discharged through a set of
.{ "\y l HEPA particulate filters to reduce the amount of radioactive particulates being O
E rc: eased to the environment. This filtration of the Containment Building
"[
ok Ventilation System exhaust implements the requirements of 10 CFR Part 50.36a, M
$ General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design b
) objectives givan in Section ll.D of Appendix l to 10 CFR Part 50.
, s l i $ With Containment Building Ventilation System exhaust being dischargod
$ k I JN without filtration, prepare and submit to the Commission within 30 days a Special s f 5 Report which discusses the circumstances and what action will be taken to 2
- i prevent a recurrence.
. 8
% G5 o e "
' 3.2 Stack Effluent Samplina and Analyses 5
3 5 The radioacdve gaseous discharge from LACBWR will be sampled and jf 3 g E analyzed as per the following table. 1 E N EE a- 5' l I ODCM 45 Rev. 4 I
b \; g
- G Table 3.3
- N RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS
=
j g Release .: . Sampling Minimum Analysis- . . . Type of 1 . 4 [ Type? Frequency - Frequencyf . Activity Analysis ?): 3 a Principal Gamma Emitters M l I
} E Stack Effluent Continuous M Weekly (') Particulate Sample i g q Continuous (b) Quarte @ PMimlate Sagle Sr-90 g Composite
.o '\
[o ' Z Continuous # Quarterly Particulate Sample Gross Alpha e % Composite 5 if \ji Continuous # Noble Gas Monitor Woble Gases Gross Beta 8L and Gamma o% Monthly Monthly H-3 M o Q NOTES: v (a) The filter sample shall be changed at least weekly, and filter analyses shall be
, s comrMed within seven (7) days.
.o s
' (b) The ratio of the sample flow rate to the sampled stream flow rate shall be known for Bi E the time period covered by each dose or dose rate calculation,
.0 %
"x '
(c) The principal gamma emitters for which the LLD specification applies exclusively D j are the following radionuclides: Mn-54, Co-60, Zn-65, Cs-134, Cs-137, and Ce-144 g for particulate emissions. This list does not mean that only these nuclides are to be y , considered. Other gamma peaks that are identifiable and measurable, together with those of the above nuclides, shall also be analyzed and reported in the annual Radioactive Effluent Release Report, j $ (d) When upper cavity is flooded or FUEL HANDLING is being performed, stack tritium O grab samples will be taken at least once per seven (7) days. x (e) Lower Limits of Detection (LLD) are determined in accordance with plant
$ procedures and are calculated in accordance with criteria of NUREG-0473,
.E o Revision 2.
E $ TO E v a . a $ O-- ODCM 46 Rev. 4
3.3.3 Stack Effluent Release Limitation g - a) Instantaneous Dose Rate - the dose rate due to radioactive materials O , released in gaseous stack effluents to areas beyond the EFFLUENT ; 1 E i g , RELEASE BOUNDARY shall be limited to: e l 9 The dose rate limit for noble gases shall be 5 500 mrem / year to the [ + O total bod" and 5 3000 mrem / year to the skin. 5 : p , + The dose rate limit for H-3 and for all radionuclides in particulate e$ k form with half-lives greater than 8 days shall be 51500 mrem / year to
- g. any organ.
o 4 y Q R The dose rate due to noble gases in gaseous stack effluents shall be
- determined to be within the above limits in accordance with Section 2.0, f3 N
g Offsite Dose Calculations. 5.. D x' The dose rate due to H-3 and for all radioactive materials in E g particulate form with half-lives > 8 days in gaseous stack effluents shall 5 R s be determined to be within the above limits in accordance with e Section 2.0, Offsite Dose Calculations, by obtaining representr'ive 5 $ 0 samples and performing analyses in accordance with the sampling and 5 analysis program specified in Table 3.3.
? z sc) 8a
'g y if the dose rate (s) exceeds the above limits, without delay decrease v i
$ y the release rate to within the above limit (s). l gx 1 1
&5 f} -
/
ODCM 47 Rev. 4
{ y_ This instantaneous dose rate limit is provided to ensure that the g\ dose rate at any time at the EFFLUENT RELEASE BOUNDARY from o- q J gaseous effluents from LACBWR will be within the annual dose limits of
=~
h 10 CFR Part 20 for unrestricted areas. The annual dose limits are the
}lE '
doses associated with the concentrations of 10 CFR Part 20, Appendix B, 23 j Table 2, Column 1. These limits provide reasonable assurance that 8 i E , radioactive material discharged in gaseous effluents will not result in the
$ (
j @ exposure of an individual in an unrestricted area, outside the EFFLUENT o k - RELEASE BOUNDARY to annual average concentrations exceeding the
% limits specified in Appendix B, Table 2 of 10 CFR Part 20. For individuals 5 N O
( who may at times be within the EFFLUENT RELEASE BOUNDARY, the g . occupancy of the individual will be sufficiently low to compensate for any i s increase in the: atmospheric diffusion factor above that for the EFFLUENT ! s x. RELEASE BOUNDARY. The specified release rate limits restrict, at all l 5 g times, the corresponding gamma and beta dose rates above background a l I
- to an individual at or beyond the EFFLUENT RELEASE BOUNDARY to 5 500 mrem / year to the total body or to 5 3000 mrem / year to the skin.
E 2 N These release rate limits also restrict, at all times, the corresponding 5 organ dose rate above background to an infant via the inhalation pathway
? z
.e o i
? $ to 51500 mrem / year. l 5 E S i t 5 O
ODCM 48 Rev. 4 l
g_ __ b) Dose from Noble Gas - the air dose to a MEMBER OF THE PUBLIC gk due to noble gases released in gaseous effluents to areas beyond the o EFFLUENT RELEASE BOUNDARY shall be limited to: Calendar Quarter E E 55 mrad from gamma radiation o( g 510 mrad from beh particulate radiation I x - E \ Calendar Year 56 o gg 510 mrad from gamma radiation
$ 20 mrad from beta particle radiation v $
N 16 O The cumulative dose contributions shall be determined at least once per O V g A calendar quarter in accordance with Section 2.0, Offsite Dose N s 2 { w . Calculations. 2 With the calculated air dose from radioactive noble gases in i
$. s gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, a Special Report which identifies the o g cause(s) for exceeding the limit (s) and defines the corrective actions g which have been taken or will be taken to reduce the releases of E z s 8 radioactive noble gases in gaseous effluents so that the cumulative dose e d 1 z during each subsequent quarter and the dose for the calendar year will be j j g within the above limits.
n L 5 \ 0.] \ ODCM 49 Rev. 4 i
i g__ This limit is provided to implement the requirements of Sections 11.B, g ; lil.A, and IV.A of Appendix 1,10 CFR Part 50. The dose calculations in Q~ l the ODCM implement the requirements in Section Ill.A of Appendix 1 that h conformance with the guides of Appendix I is to be shown by calculational e
}l procedures based on models and data such that the actual exposure of E s' O j an individual through the appropriate pathways is unlikely to be E ! substantially underestimated. 1 x ,
E \ . b c) Dose from Radionuclides other than Noble Gases - the dose to a o
==
k MEMBER OF THE PUBLIC from H-3, and all radionuclides in particulate {$ form with half-lives greater than 8 days, in gaseous effluents released to R F areas beyond the EFFLUENT RELEASE BOUNDARY shall be limited to: y h
.g g Calendar Quarter 5 7.5 mrem to any organ R
s 5 Calendar Year 8 I w 515 mrem to any organ e
~
The cumulative dose contributions shall be determined at least once per E
-y calendar quarter in accordance with Section 2.0, Offsite Dose m
], 8 Calculations.
E $ O E . E b a 5
/3--
LJ ODCM 50 Rev. 4
h g__ With the calculated dose from the release of H-3 and al! g , radionuclides in particulate form with half-lives greater than 8 days, in
- O' gaseous effluents exceeding any of the above limits, prepare and submit a r g to the Commission within 30 days a Special Report which identifies the iE
- cause(s) for exceeding the limit and defines the corrective actions which 1 E
] O , have been taken or will be taken to reduce these releases in gaseous T e effluents during remaining quarters so that the cumulative dose during m j each subsequent quarter and during the calendar year will be within the e k 2 h- > above limits. 4 i j ( This limit is provided to implement the requirements of Sections ll.C, s , _Y lil.A, IV.A and Annex of Appendix 1,10 CFR Part 50. The ODCM n . kJ g.
' calculational methods specified in the surveillance requirements l
'5 t 5 implement the requirements in Section Ill.A of Appendix I that fg % conformance with the guides of Appendix I be shown by calculational l
procedures based on models and data such that the actual exposure of I , an individual through appropriate pathways is unlikely to be substantially e j $ underestimated. R E I 8s 5 E J E hk - t s O ODCM 51 Rev. 4 J
7._. l i ! 3.3.4 Instrumentation l i b% i The radioactive gaseous effluent monitoring instrumentation channels o shown in Table 3.4 shall be OPERABLE with their alarm and/or trip setpoints set < , ia j . to ensure that the limits of Section 3.3.3a are not exceeded. The stack noble i jE gas instrumentation alarm setpoint will be determined and adjusted in i I Oj accordance with the methodology and parameters in Section 2.0, Offsite Dose ! $ N 3 z - Calculations. E ( l
- 8 i E t - The radioactive gaseous effluent instrumentation is provided to monitor o -
i == and control, as applicable, the releases of radioactive materials in gaseous l a o
@ Rs effluents during actual or potential releases of gaseous effluents. The only significant noble gas remaining is Kr-85. The alarm setpoints for these e g N instruments shall be set to ensure that the alarm will occur prior to exceeding the ,
1 j l 5 limits of 10 CFR Part 20. , . h% 4 2 a i
- g a) Gaseous Effluent Instrumentation Surveillance Reauirements -
4- %
* "* Each radioactive gaseous effluent monitoring instrumentation channel l
- e shall be demonstrated OPERABLE by performance of the CHANNEL i $ $
j CHECK, SOURCE CHECK, CHANNEL FUNCTIONAL TEST, and 5 CHANNEL CAllBRATION operations at the frequencies shown in 1 8 8 m, Table 3.5. E z o . D 4
& 5 O
ODCM 52 Rev. 4
I l q b) Corrective Action , g _M ' S\ With a radioactive gaseous effluent monitoring instrumenta-ON\ (1) i 4 y tion channsi alarm and/or trip setpoint less conservative than that e R ( required, declare the channel inoperable or change the setpoint so 9
$l that it is acceptably conservative.
Oh s .
'5 e
i-(2) With less than the minimum number of radioactive gaseous e
.$ effluent monitoring instrumentation channels OPERABLE, take the E&
h% ACTION required by Table 3.4. Exert best efforts to return the , sa instruments to OPERABLE status within 30 days and, if un-0
$ successful, explain in the next Annual Radioactive Effluent
( Release Report why the inoperability was not corrected in a timely O .o
, x s
4 g s manner, x
*m r 2
a. fi 8 I s e 8 w as U N 5
? z
.e o o E e a.
l es o L.I ODCM 53 Rev. 4
.l
i Table 3.4 i RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION l 8 MINIMUM INSTRUMENT CHANNELS APPLICABLE y OPERABLE CONDITIONS ACTION ! k 4
- 1. Reactor Containment Building
. Ventilation Monitor System l E
Q n
- a. Particulate Activity Monitor 1 A l j b. Gaseous Activity Monitor 1 A 5
E y c. Sampler Flow Rate Measuring 1
-B ;
D Device 8_ % h O
- 2. Stack Monitor System
- a. Gaseous Activity Monitor 1
.C
- b. Particulate Activity Monitor 1 D ,
E Q y c. Sampler Flow Rate Measuring Device 1 B s
- When Containment Building Ventilation System is in operation. ,
.h '. s " At all times, unless alternate monitoring is available su '
a ACTIONS.: N I A. With the number of channels OPERABLE less than required by the Minimum il Channels OPERABLE requirement, effluent releases through this pathway may
$ s continue as long as a stack monitor is OPERABLE; otherwise, secure the Containment Building Ventilation.
g B. With the number of channels OPERABLE less.than required'by the Minimum E 3 Channels OPERABLE requirement, effluent releases via this pathway may continue O provided the flow rate is estimated at least once per 24 hours. g C. With the number of channels OPERABLE less than required by the Minimum z Channels OPERABLE requirement, effluent releases via this pathway may continue ! 1B j O provided the Containment Building Gaseous Activity Monitor is OPERABLE; j y d otherwise, secure the Containment Building Ventilation.
- z
_i D. With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue y @E provided continuous collection of samples with auxiliary sampling equipment is initiated within 12 hours. O ODCM 54 Rev. 4
l i I Table 3.5 ^ G1,, . RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION i 5 . SURVEILLANCE REQUIREMENTS l l l r- CHANNEL ! $ CHANNEL SOURCE FUNCTIONAL CHANNEL M I INSTRUMENT CHEECK CHECK TEST CAllBRATION j
}e s
- 1. Reactor Containment Building Ventilation Monitor System i
O DAILY MONTHLY QUARTERLY AT LEAST ONCE g a. Particulate Activity Monitor W
- PER 18 MONTHS
-{
x c l $ b. Gaseous Activity Monitor DAILY MONTHLY QUARTERLY AT LEAST ONCE W PER 18 MONTHS j D c. Sampler Flow Rate DAILY MONTHLY QUARTERLY AT LEAST ONCE W PER 18 MONTHS j o k Measuring Device ] , g .
- 2. Stack Monitor System
- a. Noble Gas Activity Monitor DAILY MONTHLY QUARTERLY AT LEAST ONCE
$*' W PER 18 MONTHS s b. Particulate Activity Monitor DAILY !"A QUARTERLY AT LEAST ONCE g W PER 18 MONTHS
.g s
%y DAILY IA QUARTERLY AT LEAST ONCE
- c. Sampler Flow Rate Measuring Device W PER 18 MONTHS h.
S NOTES: E g (1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of I %. this pathway, and control room alarm annunciation occurs if any of the following conditions exist:
- a. Instrument indicates measured levels at or above the alarm setpoint.
s $ b. Instrument indicates a downscale failure (provides control room annunciation alarm j
$ $ only).
- c. Instrument indicates a circuit failure (provides control room annunciation alarm only).
]
Ei (2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm
? z annunciation occurs if any of the following conditions exist:
y @ a. Instrument indicates measured level above the alarm setpoint on one channel.
$- [Ij b. Instrument indicates a failure by a Low Flow and Low Count Rate signal.
z
$ .J (3) The CHANNEL FUNCTIONAL TEST shall also demonstrate that the control room local
$ $ alarm occurs if the flow instrument indicates measured levels below the minimum and/or y% above the maximum alarm setpoint.
- a. 5 The CHANNEL CALIBRATION shall be conducted in accordance with plant procedures.
O-- (4) ODCM 55 Rev. 4
i I 3.4 Total Dose to a Member of the Public 5 ; Q . The dose equivalent to any MEMBER OF THE PUBLIC due to release of l l radioactivity and radiation, shall be limited to 5 25 mrem to the total body or any organ
% i
( , (except the thyroid, which is limited to 5 75 mrem) over a period of one calendar year. E d g With the calculated doses from the release of radioactive materials in liquid or o
.! gaseous effluents exceeding twice the calendar year dose limits specified in Sections
& s y 3.2.2b, 3.3.Fb, or 3.3.3c, a determination should be made, including direct radiation e i j from reactor containment and radioactive waste storage tanks to determine if the above limits have been exceeded. If these limits have been exceeded, prepare and submit a j o
S S Special Report (including an analysis which estimates the radiation exposure to a O $ MEMBER OF THE PUBLIC for the calendar year) to the Director, Nuclear Reactor A
,$ Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555, within 8
jg 30 days, which defines the corrective action to be taker, to reduce subsequent releases f to prevent recurrence of exceeding these limits. if the release condition resulting in the E j excess has not already been corrected, the Special Report shall include a request for a s variance in accordance with the provisions of 40 CFR 190. Submittal of the Special o 8 g Report is considered a timely request, and a variance is granted until staff action on the
, request is complete.
co E z
$ 8 e a p; z
- u a
- & 5 e
- V ODCM 56 Rev. 4
Cumulative dose contributions from liquid and gaseous effluents shall be 4 g determined quarterly and annually in accordance with Section 2.0, Offsite Dose Q . Calculations. y E p Cumulative dose contributions from direct radiation from the reactor containment or radioactive waste storage tanks shall be determined once per year in accordance hf 0 t
,R 6 with Section 4.0, Radiological Environmental Monitoring Program, p-,
j This requirement is provided to me,et the dose limitations of 40 CFR 190. 6N Whenever the calculated doses from plant radioactive effluents exceed twice the h design objective doses of Appendix 1, a Special Report will be submitted which describes a course of action which should result in the limitation of dose to a real g individual for 12 consecutive months to within the 90 CFR 190 limits. .
'!E :::
For conservatism, for compliance with this limit, the maximum total dose to any M% [ MEMBER OF THE PUBLIC will be assumed to be the sum of the maximums from each E a dose pathway even though the actual maximally exposed individual for each of the I %
------ pathways could not be the same person.
- s. E -
E The maximum potential dose to a MEMBER OF THE PUBLIC from direct 5 radiation from the containment and radioactive waste storage tanks is determined by 3 8
~5 $ TLD dosimeters located at various locations around the perimeter of the LACBWR 5 E
$ i ax n
U ODCM 57 Rev.4
q _ access controlled area and the EFFLUENT RELEASE BOUNDARY for the
\ ,
g environmental monitoring program. Fci compliance with this limit, the actual maximum O : possible exposure to an actual MEMBER OF THE PUBLIC from direct radiation may be n ,*
$ determined from maximum possible exposure times relative to the continuous exposure e }
dose measured by the TLD's. Conservative maximum possible exposure times will be
].
E 8 i determined by actual observation of the areas of interest by LACBWR management L 3 and/or security personnel. a: 8 & e E O R . G, s
, z e
B % i ee s 8 8 0 n E z i 8 e d B z 2 " 1 ih 2 5 O ODCM 58 Rev. 4 1
I H- 4.0 RADIOLOGICAL ENVIRONMENTAL MONITORING 5 PROGRAM Bu j 4.1 Proaram Requirements E a ,
)E The Radiological Environmental Monitoring Program (REMP) shall conform to O the guidance of Appendix I to 10 CFR Part 50. The REMP shall provide the require-1 ,
ments for monitoring, sampling, analyzing, and reporting radiation and radionuclides in j ,(' m
.6 the environment resulting from the LACBWR facility and/or its effluents. These y@
h requirements have been established to ensure the measurements of radiation and of a radioactive material in potential exposure pathways to MEMBERS OF THE PUBLIC are O ( t performed. Various environmental samples will be taken within the area surrounding
~
LACBWR and in selected controlled or background locations. An interlaboratory e s Comparison Program shall be established to ensure that independent checks on the
]
E f precision and accuracy of the measurements of radioactive materialin the l g environmental sample matrices are performed as part of the Quality Assurance l
~
l
- Program for environmental monitoring.
1 e jk Pi The radiological monitoring program required by this specification provides measurements of radiation and of radioactive materials in those exposure pathways
] y$
S and for those radionuclides, which lead to the highest potential radiation exposures of jE individuals resulting from plant effluents. This monitoring program theory supplements l u J
$ $ the radiological effluent monitoring program by verifying that the measurable ae fm concentrations of radioactive materials and levels of radiation are not higher than U l ODCM 59 Rev. 4
l-q expected on the basis of the effluent measurements and modeling of the environmental g O y ' exposure pathways. 3 The requirement for participation in an Interlaboratory Comparison Program is e e provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental samples are performed to O
,[ demonstrate that the results are reasonably valid.
R dN 8 i 4 4.2 REMP Description h% Radiological environmental monitoring samples will be collected and analyzed in e E 3. accordance with Table 4.1. The specific sample locations are listed in HSP-03.1. t C,_D , Section 3 of the Health and Safety Procedures (HSP
.e ' provide specific guidance to the HP technicians in the collection and analysis of each B 4
$ environmental sample.
1N 5 8 I,
- 5 $
o R E E z s8 e a B Z
? #
EE a 5 C U ODCM 60 Rev. 4
O Date H:alth Physics Review G Data Operations Rnview Comm. Approvil Date T l Prepared or Revised By l LARRY L. NELSON 3/25/96 M w m f( M y//i/9/,,
/s C (34Wh o/dv/ l
" Table 4.1 .
RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Number of Sampling and Collection and/or Sample Samples W Frequency Type and Frequency of Analysis
- 1. AIRBORNE Three (3) Continuous operation (1) Analyze each filter for gross beta PARTICULATES of sampler with radioactivity > 24 hours following filter sample collection as change. Perform gamma isotopic analysis required by dust on each sample when gross beta activity is loading, but at least > 10 times the control sample (La Crosse).
weekly. (2) A composite of particulate filters from each location will be gamma analyzed at least once per quarter.
- 2. DIRECT Eight (8) - At least 1) Gamma dose - at least semi-annually.
RADIATION at least 2 monitors semiannually. at each location.
- 3. WATERBORNE Two (2) Monthly. 1) Gamma isotopic analysis monthly on each (River Water) sample.
- 2) Tritium analysis on composite sample from each location quarterly.
(A) Exact sample locations are listed in HSP-03.1. ODCM 61 Rev.4
O Prepared or Revised By Date H:alth Physics Review O Data Operations Review Comm, Approval Data el
\ LARRY L. NELSON 3/25196 kmm / (g4//j&
'f/II/9(s f}T ff:u&R g///p.g l
# Ta61e '4.1 - (cont'd)
RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Number of Sampling and and/or Sample Samples W Collection Frequency Typa and Frequency of Analysis
- 4. RIVER Two (2) Semi-annually. (1) Gamma isotopic analysis on each SEDIMENT sample.
- 5. INGESTION
- a. Fish One (1) sample of two At least semi-annually. (1) Gamma isotopic analysis of the (2) different species in edible portions of each sample.
area important as a recreational or commercial species.
- b. Milk One (1) At least monthly when (1) Gamma isotopic analysis on each animalc are on pasture sample.
(May thru October).
- c. Vegetation One (1) At time of harvest. (1) Gamma isotopic analysis of the edible portion of each composite sample.
(A) Exact sample locations are listed in HSP-03.1. ODCM 62 Rev.4
- -. -. - =- .- .- . - . . - . - - . - .
J 1 ~ 4.3 REMP Lower Limits of Detection (LLD) l F-y . The sampling techniques and counting equipment used for the analysis of i O samples collected as requirements of the REMP will meet LLD's calculated in 2 e accordance with criteria of NUREG-0473, Rev. 2. LACBWR's LLD's are calculated R
<l g as follows and are essentially the same as those found in NUREG-0473, Rev. 2.
b Table 4.2 lists these values. 3 x 8 ' i e e . g - o ' o-4 l9 ia t
~
l
$N ,
2 a. r is I w , i $ o R 5 1B z s 8 e e b Z - e a. e s e O ODCM 63 Rev. 4
Prepared or Revised By Date Health Physics Review Date Operations Review Comm. Aoproval Date, t.ARRY L. NELSON 3/25/96 Mm% / Mg Y////9/,
/(/7-[Jw h dj'/9Y f / // /
4.3.1 Calcuiation of Lower Limits of Detection: The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95% probability, with 5% probability of falsely concluding that a blank observation represents a "real" signal. For a particular measurement system (which may include radiochemical separation): 4.66 S3 LLD = E x V x 2.22 x Y x Exp (-AAt) WHERE: LLD is the priori lower limit of detection as defined above (as picccurie per unit mass or volume). S, is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute). E is the counting efficiency (as counts per gamma). V is the sample size (in units of mass or volume). 2.22 is the number of transformations per minute per picocurie. Y is gamma abundance for isotope of interest. A is the radioactive decay constant for the particular radionuclide. At is tha elapsed time between sample collection (or end of the sample collection period) and time of counting. ODCM 64 Rev.4
O Prepared or R:: vised By Date _ Health Physics Review e Data Operations Revipw Ccmm. Approv 1 D2t3 7 LARRY L. NELSON 3/25/96 he6MM '/////f(,
~
g[C @wMh' c////# g l f / // / Table 4.2 ENVIRONMENTAL SAMPLE ANALYSES LOWER LIMITS OF DETECTION VALUES (LLD) Sample Type Water Airbome Particulate Fish Milk Sediment Analysis pCill or Radiciodine (pCi/m ) (pCi/Kg, Wet) (pCill) (pCi/Kg Dry) Gross Beta 6 1 E-2 H-3 3500(2000)' Mn-54 15 130 Cos60 15 130 - Zn-65 30 260 Cs-134 15 5 E-2 130 15 150 Cs-137 18 6 E-2 150 18 180
- For drinking water.
t ODCM 65 Rev. 4 ,
4.4 Interlaboratory Comparison Proaram
- 5. ,
n An Interlaboratory Comparison Program will be established to ensure that the n O : , analyses being performed to comply with the REMP is accurate. A suitable offsite )
- - 1 y laboratory will be used to supply NIST traceable or equivalent standard spiked sample
& \
5 media for analysis. The offsite laboratory will supply a rel port to DPC of the comparison 1 E' i E s 1 gj Jg results. The Interlaboratory Comparison Program will be conducted annually. The results of this comparison will be included in the Annual Radiological Environmental l 6 j Monitoring Report. W ' i o ! 4.5 Reportina Reauirements . o Q I An Annual Radiological Environmental Monitoriig Report shall be g a) g__ submitted to the Administrator of the Regional Office of the NRC. This report i ; shall include summarized and tabulated results, including interpretations and a: h i analysis of data trends, of environmental samples taken during the previous g calendar year, in the event that some results are not available for inclusion with 8
* % the report, the report shall be submitted noting and explaining the reasons for
, g the missing results. The missing data shall be submitted as soon as possible in 15 2 E a supplementary report.
A The report shall also include the following: a summary description of the E z
.e o
!; $ Radiological Environmental Monitoring Program, a map of all sampling locations E E
$ J keyed to a table giving distances and directions from the plant, the results of the e s e)
L ODCM 66 Rev. 4
Interlaboratory Comparison Program, and a discussion of all analyses in which ( P -(
. the LLD was not achievable.
16 o-b) With the REMP program not being conducted as specified in Table 4.1, 5: # l [ i prepare and submit to the Commission, in the Annual Radiological Environmental Monitoring Report, a description of the reasons for not conducting l I the program as required, analysis of the cause of unexpected results, and the a: , plans for preventing a recurrence. E
.9 E
& c) With the Interlaboratory Comparisons not being performed, report the o
corrective actions taken to prevent a recurrence to the Commission in the n 2 3 b- Radiological Ens ironmental Monitoring Report.
%v h 's N
t d) With radiological environmental sample analysis in excess of the
& reporting levels listed in Table 4.3, when averaged over any calendar quarter,
.0 N
} g prepare and submit to the Commission a Special Report within 30 days, with a
- a. y
$ description of the reasons for exceeding these reporting levels.
E s s E o O n A E z i 8 e a o Z v a
& 5 9 67 Rev.4 ODCM
l l g-- Table 4.3 i k Reporting Levels for Radioactivity Concentrations l
.n in Environmental Samples l E
e E - W ateri Airbome' Particulate Fishi: Milk.- 4 g . Analysis - i.pCille . pCi/m' ~ p Ci/kg (wet) pCill l E u 0 H-3 20,000 - -- - E
'f g Mn-54 1,000 -
30,000 -- f,. g Co-60 300 - 10,000 h ; Zn-65 300 - 20,000 - g Cs-134 30 10 1,000 60 o b g, Cs-137 50 20 2,000 70 C: s ! s
.B i x
\
" i e
I s
- E 3 5 e s IE v i
! 5 O
ODCM 68 Rev.4
1
LA CROSSE BOILING WATER REACTOR PROCESS CONTROL PROGRAM (PCP) l l
Prepared br[ l 9 6
/~ Health Physics Review:~ Y [7b V I // / Date QA Review: bm -
v-*/-1'4 ORC Approved: M / 9[ 6att 1 I March 1996 (Revision 5) Dairyland Power Cooperative 3200 East Avenue South La Crosse, WI 54602 O I l
(- U Safety Analysis in Accordance with Decommissionina Plan and Technical Specifications: (1) Will the probability of occurrence or the consequences of an accident or malfunction of equipment necessary for SAFSTOR previously evaluated either in the Decommissioning Plan or FSAR be increased? NO 1 Justification For Answer: The removal of a Tech. Spec. reference number and reorganizing of the I PCP will not alter the PCP to increase the probability of an accident or i malfunction important to safety previously evaluated. l (2) is there a possibility of an accident or malfunction of a different type than previously evaluated in the Decommissioning Plan or FSAR being created? NO Justification For Ansv,er: The handling of radioactive waste will not be changed in our Radiation Protection Program or PCP. Only the Tech. Spec. reference has been removed. A U (3) is the margin of safety as defined in the bases for any. Technical . Specifications reduced? NO ) Justification For Answer: 1 The Tech. Spec. criteria has not changed for the PCP. There is no program change from that specified in Technical Specifications. (4) Will the proposed change result in a significant environmental impact not previously evaluated in the Environmental Assessment in support of the August 7,1991, Decommissioning Order or the Final Environmental Statement 4 ES) related to operation of LACBWR, dated April 21,1980 (N U R EG-0191)? NO Justification For Answer: The. Tech. Spec. criteria is unchanged, so the PCP basis is unchanged. The PCP will continue to meet all criteria in order to ensure the safe handling l storage and shipment of radioactive waste. O V PCP Rev.5
O LACBWR kJ PROCESS CONTROL PROGRAM (PCP) I. INTRODUCTION The purpose of the Process Control Program (PCP) is to describe the methodology and procedures used by the La Crosse Boiling Water Reactor (LACBWR) Facility during SAFSTOR, to ensure that radioactive waste material processing, packaging, transportation and disposal will be accomplished in a manner that is in compliance with all applicable federal, state, and burial si,te regulations. The PCP will provide guidance for LACBWR's two basic waste streams: (1) dry active waste (DAW), or (2) wet radioactive waste (resin or filter media). II. SPECIFICATION Technical Specifications state; l O a. "The PCP shall be maintained onsite and will be available for NRC 1 review." !
- b. " Licensee-initiated changes to the PCP shall be submitted to the Commission in the annual Radioactive Effluent Release Report for the period in which the change (s) was m' ade. This submittal shb.1 contain:
Information to support the rationale for the change; a A determination that the change did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and
- Documentation of the fact that the change has been reviewed and found acceptable by the ORC."
O PCP 1 Rev.5
Ill. WET RADIOACTIVE WASTE The PCP and Health and Safety Procedure (HSP) 4.1 establish the steps used to dewater and stabilize wet radioactive waste material from the process water treatment systems and the liquid waste system. These wastes will be ! transferred to the Waste Treatment Building (WTB) for storage and processing before its ultimate disposal. The WTB is located to the ' northeast of the Containment Building. The WTB l j contains f acilities and equipment for the collection, processing, storage and handling of low level solid radioactive waste materials. The grade floor of the WTB contains a shielded compartment which l : encloses a permanent 320 f t3 stainless steel Spent Resin Receiving Tank (SRRT) !' with associated resin receiving and transler piping, SRRT recirculation piping, SRRT demineralized water flush piping, SRRT dewatering piping, SRRT air .cparger piping, valves, and a SRRT ultrasonic levelindication assembly. This i cubicia is in the southeast corner of the building and has nominal 2'6" to 3' thick reinforud high density concrete shield walls on the outer sides. Wet radioactive waste materials may typically be described as whole bead demineralizer resins, mechanical cartridge filters, waste sludge, and inorganic l particulate residues. When expended, these wet wastes are transferred to the SRRT or, in some cases, directly into an HIC for processing. In the case of l Class C wet radioactive waste, as an alternate method to transfer to the SRRT, they may be transferred to a HIC inside a transportation ci:sk and dewatered from the cask to the plant process liquid waste system, thus reducing radiation exposure to personnelinvolved in performing the necessary package handling ; operations. Samples of wet waste are analyzed for isotopic composition and for I waste classification.
)
Expended demineralizer resins and waste sludge material from sumps and tanks are transferred to the SRRT as described in the Operation Manual Volume Vll. . l From the SRRT, partially dewatered spent resins and/or sludge material, which have been sluiced from plant demineralizers and/or tanks, are pumped at a low flow rate / pressure with a Roper-type screw pump to a HIC, located in the adjacent shielded cubicle as described in Operations Manual Volume Vll. The wall between the SRRT and the disposal HIC and other walls surrounding the HIC are a nominal 2' thick high density concrete. Also located in this cubicle is an air-driven Sandpiper-type diaphragm HIC dewatering pump, the resin transfer and dewatering valves and hoses, and the disposal HIC level indication connecting cable. Once in the HlC, spent resins and sludge material is dewatered to the SRRT which is PCP 2 Rev.5
1 I O subsequently dewatered through a dewatering ion exchanger (DWlX' A final O dewatering is performed as described in Operations Manual Volume /ll and Health l
- and Safety Procedure (HSP) 4.1.
The final dewatering will ensure that the contents of the HIC meets the requirements of 10CFR61.56(a)(3) and 10CFR61.56(b)(2) and burial site criteria ; for free-standing liquids. Stabilization of the dewatered wet waste is provided by the HIC as authorized by 10CFR61.56(b)(1). The contractor which supplies the - HlC will. provide DPC with a copy of the Certificate of Compliance for the HIC j which details specific limitations on the use of the HIC. J Once final dewatering is completed, the HlC is inspected, sampled, and the HIC lid is sealed and closed. The HIC is surveyed, decontaminated, and loaded into an appropriate shipping cask. Dewatered spent mechanical cartridge filters may be added directly to a HIC containing spent resins, if necessary. 1 j The shipping cask is prepared for shipment in accordance to contractor's i procedures. The vehicle is inspected both upon arrival onsite and after it is loaded prior to departure in accordance with procedure. ; 4 Once properly prepared for shipment, the wet waste will be either sent directly to an approved disposal site for burial, or it will be sent to a licensed waste
- processor for volume reduction. Any material remaining after reprocessing will be i sent to burial by the processor for DPC. l
\
i IV. DRY ACTIVE WASTES The PCP and Health and Safety Procedure 4.2 establish the steps used to I process, package and ship the Dry Active Wastes (DAW) from the plant, in j accordance with 10CFR20.2006. The Dry Active Waste materials may typically be i described as paper, cloth, metal, wood, plastic, concrete and other items or components which have become contaminated with radioactive materials. The DAW is normally characterized and labeled as Class A Unstable in accordance with 10CFR61.55 and .56. l DAW will either be processed at LACBWR for burial at an approved burial site, or it will be processed for shipment to a licensed reprocessing contractor.
- a. DAW for direct burial These wastes processed for burial from LACBWR are normally compacted using a DAW compactor unit located in the WTB. They are packaged in DOT Spec 7A containers or Spec 17H drums. Sufficient PCP 3 Rev.5
._.__.__._--.._.___._._.._____.-.-____...___m _ . . ~ _ . . _
l I l l absorbent materialis added to each waste container to limit free
'O standing and non corrosive liquid to less than 1% of the volume. Some
! dry waste material may be plant components which had become i l activated by neutron irradiation. Representative samples of DAW are analyzed for isotopic composition and for waste classification. After packaging, these DAW wastes are transferred to the DAW storage ! I j building which is located southwest of the Turbine Building. It is used to store processed, packaged and sealed low level dry active waste materials, and sealed low level activity components until they are sent to burial. The building has the capacity for 500 DOT 17H-55 gallon drums of waste.
- b. DAW for Reprocessina Containers supplied by the contractor will be strategically placed within the controlled area to allow for the most efficient loading of these containers. The container size will vary due to the application required, Once filled, the reprocessing contractor will be notified and shipping arrangements will be made. The containers will be sealed and surveyed ,
L before leaving LACBWR. Any waste volume remaining after I reprocessing will be sent to burial by the contractor for LACBWR. V. WASTE CI,ASSIFICATION LACBWR wastes will be classified as Class A Stable or Unstable, Class B Stable, cr Class C Stable to determine the acceptability for near-surface disposal and for the purpose of segregation at the disnosal site. The waste class will be l based on the concentration of certain radionuclides in the waste as outlined in j 10C FR61.55. Radidnuclide concentrations will be determined based on the volume or weight of the final waste form as discussed in Section C.2 of the Branch Technical l Position Paper on Waste Classification. l Of the four suggested methods for determining radionuclide concentration, ! the one most commonly used is the direct measurement of individual radionuclides I (gamma emitters) and the use of scaling factors to determine the radionuclide concentration of difficult to measure radionuclides (normally non-gamma emitters). The use of the other suggested methods; material accountability, classification by source or gross radioactivity measurements may occur if the situation best fits the ! use of that methodology. !O PCP 4 Rev.5
l l Plant procedures are used in the determination of radionuclide concentra-l~ I tion for difficult to measure nuclides (normally non-gamma emitters) and for the classification of radioactive waste for near-surface burialin accordance to 10CFR61.56(a)(3) and Table 1 and Table 2. VI. SillPM ENT M ANIFEST l l Each shipment of solid radioactive waste to a licensed land disposal facility will be accompanied by a shipment manifest as required by 10CFR20,2006. The manifest will contain the name, address, and telephone number of the waste l generator. The manifest will also include the name, address, and telephone ! number or the name and EPA hazardous waste identification number of the person transporting the waste to the land disposal facility. The manifest will also indicate to the extent practicable; a physical l description of the waste; the volume; radionuclide identity and quantity; the total radioactivity; and the principal chemical form. The solidification agent, if applicable, will be identified. 1 Waste containing more than 0.1% chelating agents by weight will be identified and the weight percentage of the chelating agent estimated. Waste classification, Class A, B, or C, will be clearly indica *ed on the manifest. The total quantity of the radionuclides H-3, C-14, Tc-99, and 1-129 will be shown on the manifest. Source material mass, and the masses of U-233, U-235, Pu isotopes and total Special Nuclear Material will be calculated and shown on the manifest. I Each manifest will include a certification by DPC that the transported materials are properly classified, described, packaged, marked, and labeled, and are in proper condition for transportation according to the applicable regulations of the Department of Transportation and the NRC. A qualified individual will sign and date the manifest. LACBWR procedures are used for the preparation of burial site radioactive shipping manifests. LACBWR will maintain a manifest record-keeping and tracking system that meets the requirements of 10CFR20.2006. l Each shipment of radioactive waste to a licensed reprocessing contractor
- will be accompanied by s shipment manifest that meets the requirements of that i i contractor's license.
i i i i: O ; ! l PCP 5 Rev.5 l l l
. . . . - - - - . . . - . ~ - . - . . - - . - ~ ~ - - - . ~ . = - . . - - . . . . . - . . - .
l. i ! VIII. ADMINISTRATIVE CONTROI,S
- A. Trainino t
A training program will be conducted to ensure that waste processing ! will be performed according to plant procedures and the PCP. An ] individual's training record will be maintained for audit and inspection. ; The processing and shipment of radioactive material will be performed l by qualified and trained personnel. l 1
- 8. Record Retention Records of processing data, test and analysis results, results of training, inspections and audits will be retained in accordance with LACBWR Quality Assurance requirements for record retention.
I C. Documentation Control i DPC-initiated changes may be made to the PCP and procedures in t accordance with Technical Specifications and shall become effective l l upon review and acceptance by the Operations Review Committee (ORC). O Radioactive waste that does not fall within previous waste processing l. experience will be evaluated and, if necessary, included in the PCP prior ! to final processing and disposal. Approved changes in the PCP will be reported to the NRC in the subsequent annual Radioactive Effluent Release Report. D. Quality Assurance I The provisions of the NRC-approved LACBWR Quality Assurance . Program apply to all activities performed under the PCP, and solid l radioactive waste processing procedures. i E. Voluntary Information Submittal I If any of the following mishaps occur during the preparation of LLW waste for disposal, a voluntary submittal of information will be sent to the NRC. This will be a 30-day report to the NRC's Director of the Division of Low-Level Waste Management and Decommissioning and will ! also be sent to the designated State disposal-site regulatory authority. lO ( PCP 6 Rev.5
r'N. (1) Failure of the high-integrity containers used to ensure a stable d waste form. Container failure can be evidenced by changed container dimensions, cracking, or damage resulting from mishandling (e.g., dropping or impacting against another object). (2) Misuse of high-integrity containers, evidenced by a quantity of , free liquid greater than 1 percent of container volume, or by an l excessive void space within the container. Such misuse is l 1 prohibited by 10 CFR 61.56. (3) Production of a solidified Class B or C waste form that has any of the following characteristics: Contains free liquid in quantities exceeding 0.5 percent of 4 the volume of the waste. Contains waste with radionuclides in concentrations exceeding those considered dunng waste form qualification testing accepted by the regulatory agency, which could lead to errors in assessment of waste class. ; l
- Contains a significantly different waste loading than that used in qualification testing accepted by the regulatory agency.
- Contains chemicalingredients not present in qualification testing accepted by the regulatory agency, and those quantities are sufficient to unacceptably degrade the waste product.
- Shows instability evidenced by crumbling, cracking, spalling, voids, softening, disintegration, non- homogeneity, or dimensional changes.
- Evidence of processing phenomena that exceed the limiting processing conditions identified in applicable topical reports on process control plans (e.g., foaming, temperature extremes, premature or slow hardening, and production of volatile material).
PCP 7 Rev.5}}